WorldWideScience

Sample records for clad al-li alloys

  1. Retention and release of tritium in aluminum clad, Al-Li alloys

    International Nuclear Information System (INIS)

    Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the 6Li(n,α)3He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs

  2. The physical metallurgy of mechanically-alloyed, dispersion-strengthened Al-Li-Mg and Al-Li-Cu alloys

    Science.gov (United States)

    Gilman, P. S.

    1984-01-01

    Powder processing of Al-Li-Mg and Al-Li-Cu alloys by mechanical alloying (MA) is described, with a discussion of physical and mechanical properties of early experimental alloys of these compositions. The experimental samples were mechanically alloyed in a Szegvari attritor, extruded at 343 and 427 C, and some were solution-treated at 520 and 566 C and naturally, as well as artificially, aged at 170, 190, and 210 C for times of up to 1000 hours. All alloys exhibited maximum hardness after being aged at 170 C; lower hardness corresponds to the solution treatment at 566 C than to that at 520 C. A comparison with ingot metallurgy alloys of the same composition shows the MA material to be stronger and more ductile. It is also noted that properly aged MA alloys can develop a better combination of yield strength and notched toughness at lower alloying levels.

  3. Modeling-Based Processing of Al-Li Alloys for Delamination Resistance Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Al-Li alloys are of interest for use in aerospace structures due to the desirable combination of high strength and low density. However, high strength Al-Li alloys...

  4. A TECHNIQUE FOR IMPROVING THE TOUGHNESS OF Al-Li POWDER METALLURGY ALLOYS

    OpenAIRE

    Webster, D.

    1987-01-01

    A technique has been developed for increasing the toughness of Al-Li products made by powder metallurgy. The technique which involves the addition of unalloyed aluminum powder to Al-Li powder before compaction was evaluated with Al-Li-Cu-Mg-Zr alloys (Al 8090), and Al-Li-Zn-Cu-Mg-Zr and Al-Li-Mg-Si-Cr alloys . The addition of 15% aluminum to Al 8090 aged at 422K for 40 h produced an increase in impact toughness of 215% at the expense of a drop in yield strength of 11%. The Al-Li-Mg-Si-Cr allo...

  5. Effect of Impurities and Cerium on Stress Concentration Sensitivity of Al-Li Based Alloys

    Institute of Scientific and Technical Information of China (English)

    孟亮; 田丽

    2002-01-01

    A notch sensitivity factor was derived in order to evaluate the stress concentration sensitivity of Al-Li based alloys. The factor values for the Al-Li alloy sheets containing various contents of impurities and cerium addition were evaluated by determining the mechanical properties. It is found that the impurities Fe, Si, Na and K significantly enhance the stress concentration sensitivity of the alloys 2090 and 8090, whereas cerium addition reduces the stress concentration sensitivity to a certain degree for the high strength alloys. However, an excess amount of cerium addition in the high ductility alloy 1420 can significantly increase the stress concentration sensitivity. As compared with conventional aluminum alloys, the Al-Li based alloys generally show high stress concentration sensitivity. Therefore, a special attention must be paid to this problem in the practical application of Al-Li based alloys.

  6. Tensile behavior of rapidly solidified Al-Li-Zr and Al-Li-Cu-Mg-Zr alloys at 293 and 77 K

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.S. [Gyeongsang National Univ., Chinju (Korea, Republic of). Div. of Materials Science and Engineering; Shin, K.S. [Seoul National Univ. (Korea, Republic of). School for Materials Science and Engineering; Kim, N.J. [POSTECH, Pohang (Korea, Republic of). Center for Advanced Aerospace Materials

    1999-08-01

    It has been found that the tensile ductility of some Al-Li alloys increases significantly with decreasing temperature. Although this inverse temperature dependence has often been observed in some non-Li-containing Al alloys, such as Al 2219, the magnitude of improvement in tensile ductility is generally much higher in Li-containing Al alloys. Several hypothetical mechanisms have been proposed to explain the increase in tensile ductility at cryogenic temperatures for Al-Li alloys. At present, the studies on the cryogenic mechanical properties o Li-containing Al alloys are largely limited to ingot-melted alloys, and the data are not readily available for powder metallurgy (PM) processed Al-Li alloys. The refined microstructure of PM processed Al-Li alloys would minimize the extrinsic delamination effects on the tensile properties and, as a result, these may serve as better materials for studying the mechanism(s) for the improved cryogenic tensile properties in Al-Li alloys. The objective of the present study, therefore, was to examine the tensile properties of rapidly solidified (RS)/PM processed Al-Li alloys and to identify the mechanism of the increase in tensile ductility at cryogenic temperatures.

  7. Corrosion Studies of 2195 Al-Li Alloy and 2219 Al Alloy with Differing Surface Treatments

    Science.gov (United States)

    Danford, M. D.; Mendrek, M. J.

    1998-01-01

    Corrosion studies of 2195 Al-Li and 2219 Al alloys have been conducted using the scanning reference electrode technique (SRET) and the polarization resistance (PR) technique. The SRET was used to study corrosion mechanisms, while corrosion rate measurements were studied with the PR technique. Plates of Al203 blasted, soda blasted and conversion coated 2219 Al were coated with Deft primer and the corrosion rates studied with the EIS technique. Results from all of these studies are presented.

  8. Effect of aging in an electric field on microstructures and properties of 1420 Al- Li alloy

    Institute of Scientific and Technical Information of China (English)

    刘北兴; 李洪涛; 覃耀春; 冯海波

    2002-01-01

    After solution treatment, the 1420 Al- Li alloy samples were aged at different temperatures in an e-lectric field with different intensity. The measurements made showed that the electric field increased the strengthofthe 1420 Al -Li alloy, and best properties were obtained when they were aged at 120 ℃ with E = 4 kV/cmfor 12 hrs. The electric field promoted the nucleation ofδ' phase, increased the quantity of the δ' phase, andmade the size of the δ' phase particles smaller. The electric field restrained the formation and growth of PFZ,and increased the intensity of the electric field while the width of the PFZ was decreased.

  9. Corrosion Properties of Light-weight and High-strength 2195 Al-Li Alloy

    Institute of Scientific and Technical Information of China (English)

    XU Yue; WANG Xiaojing; YAN Zhaotong; LI Jiaxue

    2011-01-01

    The intergranular corrosion and exfoliation corrosion of 2195 Al-Li alloy treated by multi-step heating-rate controlled aging (MSRC)are studied.The corrosion features of 2195 Al-Li alloys which are respectively treated by high-temperature nucleation MSRC(H-M)and low-temperature nucleation MSRC(L-M)are contrasted.And the corrosion mechanism of 2195 Al-Li alloy is also discussed from the viewpoint of microstructure(types,distribution,etc.)of the strengthening phase.The results show that 2195 Al-Li alloy after H-M is more susceptible to intergranular corrosion and exfoliation corrosion than that of alloy after L-M.The degree of intergranular corrosion increases with the increase of predeformation amount and the surface parallel to the rolling direction is more prone to exfoliation corrosion.The main reason of intergranular corrosion and exfoliation corrosion is the formation of corrosion galvanic couples among T1 phase,θ' phase and grain boundary precipitate-free zones(PFZ).

  10. ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY DURING CORROSION PROCESS OF 8090 Al-Li ALLOY IN EXCO SOLUTION

    Institute of Scientific and Technical Information of China (English)

    J.F. Li; Z.Q. Zheng; C.Y. Tan; S.C. Li; Z. Zhang; J.Q. Zhang

    2004-01-01

    The corrosion behavior and electrochemical impedance spectroscopy ( EIS) features of 8090 Al-Li alloys in EXCO solution were investigated, and the EIS was simulated using an equivalent circuit. At the beginning of immersion in EXCO solution, the EIS is comprised by a depressed capacitive arc at high-mediate frequency and an inductive arc at low frequency, and the inductive component decreases and disappears with immersion time. Once exfoliation or severe pitting corrosion is produced, two capacitivearcs appear in the EIS. These two capacitive arcs are originated from the two parts of the corroded alloy surface, the original flat alloy surface and the new inter-face exposed to the aggressive EXCO solution due to the exfoliation or pitting corrosion.Some corrosion development features of 8090 Al-Li alloys in EXCO solution can be obtained through simulated EIS information.

  11. Thermal mechanic processing effects on the microstructural evolution of Al-Li alloys

    International Nuclear Information System (INIS)

    The investigation of the effects of different thermomechanical treatments on the microstructure of alloys 8090 and 8091 (Al-Li-Cu-MgZr) is the aim of the present work. In this context, the intervention of static recrystallization during solution treatment after hot working is the determining factor on the final microstructure of products in form of plates. The results could reveal that the rolling temperature is a very important variable if microstructural control is to be achieved in these alloys. (author)

  12. In Situ Assessment of Lattice in an Al-Li Alloy

    Science.gov (United States)

    Beaudoin, A. J.; Obstalecki, M.; Tayon, W.; Hernquist, M.; Mudrock, R.; Kenesei, P.; Lienert, U.

    2013-01-01

    The lattice strains of individual grains are measured in an Al-Li alloy, AA 2195, using high-energy X-ray diffraction at a synchrotron source. The diffraction of individual grains in this highly textured production alloy was isolated through use of a depth-defining aperture. It is shown that hydrostatic stress, and in turn the stress triaxiality, can vary significantly from grain to grain.

  13. Study on Damage of High Temperature Plastic Deformation for Al-Li Alloy

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The security of use for Al-Li alloy will be greatly influenced by the damage degree of plastic deformation within it at high temperature . Based on continuum damage mechanics theory, the damage evolution of Al-5.44Mg-2.15Li-0.12Zr alloy during plastic deforming at high temperature is simulated by using the damage evolution model of high temperature plastic deformation. The changing rule of its inner damage with deformation temperature, strain rate and strain is gained in this paper. The equation of damage evolution for high temperature plastic deformation is developed, providing an academic basis for the technology of plastic process of Al-Li alloys.

  14. Ultrasonic location for core end of Al-Li alloy and aluminium composite tube

    International Nuclear Information System (INIS)

    The locating method of the core end of Al-Li alloy is studied by ultrasonic wave time-frequency analysis. Adopting the high frequency and narrow pulse emission the composite metal tube can be located by means of the states of longitudinal wave and transverse wave. The dissection of the sample proves that the core end thickness is 0.1 mm and the location precision is 2 mm

  15. Superplastic deformation behavior of a rapidly solidified Al-Li alloy

    International Nuclear Information System (INIS)

    This study aims at investigating the superplastic behavior of a rapidly solidified Al-3Li-1Cu-0.5Mg-0.5Zr (mass%) alloy. Although rapidly solidified Al-Li alloys have the very fine grain structure desirable for improved superplasticity, unfavorable oxide morphology often prevents them from being superplastic. The results of superplastic deformation indicated that the proper thermo-mechanical treatment (TMT) of the alloy resulted in a much improved superplastic ductility, e.g. elongation of approximately 530%. In the case of testing at 520 C, optimum strain rate of forming was 4 x 10-2 s-1, which was one or two orders of magnitude higher than that of ingot cast Al-Li alloys. Such a high strain rate was thought to be quite advantageous for the practical application of superplastic deformation of the alloy. It could also be seen that the microstructure of the deformed alloy was similar to that of the as-received or the TMT treated alloy since continuous recrystallization was accomplished by subgrain growth and the growth of primary grains was prevented by fine β' (Al3Zr) particles. (orig.)

  16. Structure and Mechanical Properties of Al-Li Alloys as Cast

    Directory of Open Access Journals (Sweden)

    J. Augustyn-Pieniążek

    2013-04-01

    Full Text Available The high mechanical properties of the Al-Li-X alloys contribute to their increasingly broad application in aeronautics, as an alternative for the aluminium alloys, which have been used so far. The aluminium-lithium alloys have a lower specific gravity, a higher nucleation and crack spread resistance, a higher Young’s module and they characterize in a high crack resistance at lower temperatures. The aim of the research planned in this work was to design an aluminium alloy with a content of lithium and other alloy elements. The research included the creation of a laboratorial melt, the microstructure analysis with the use of light microscopy, the application of X-ray methods to identify the phases existing in the alloy, and the microhardness test.

  17. HYDROGEN EMBRITTLEMENT IN Al-Li-Cu-Mg ALLOYS (8090-T651)

    OpenAIRE

    Binsfeld, F.; Habashi, M.; Galland, J.; Fidelle, J.; Miannay, D.; Rofidal, P.

    1987-01-01

    This paper describes the hydrogen embrittlement (HE) of an Al-Li alloy aged at 190°C and with different durations of ageing (10, 15, 20 and 30 hr). Two techniques were employed to measure HE : a) cathodic polarization in a molten salts bath with -3 V/Ag on tensile specimens ; b) gaseous hydrogenation on disks. Hydrogen charging was achieved at 190°C. The results show that HE is important when the alloy is in the over-aged condition.

  18. Electronic structure and phase stability properties of Al-Li alloys

    International Nuclear Information System (INIS)

    The phase diagram of Al-Li alloys was calculated with the use of the Connolly-Williams method. In an effort to test the validity and to supplement the results of that study, equilibrium lattice constants and effective cluster interactions have been obtained using the generalized perturbation method within the first-principles multiple-scattering formalism of the Korringa-Kohn-Rostoker coherent-potential approximation. In this paper the implication of these effective interactions to the phase stability of these alloys is discussed

  19. Intrinsic fatigue crack growth rates for Al-Li-Cu-Mg alloys in vacuum

    Science.gov (United States)

    Slavik, D. C.; Blankenship, C. P., Jr.; Starke, E. A., Jr.; Gangloff, R. P.

    1993-01-01

    The influences of microstructure and deformation mode on inert environment intrinsic fatigue crack propagation were investigated for Al-Li-Cu-Mg alloys AA2090, AA8090, and X2095 compared to AA2024. The amount of coherent shearable delta-prime (Al3Li) precipitates and extent of localized planar slip deformation were reduced by composition (increased Cu/Li in X2095) and heat treatment (double aging of AA8090). Intrinsic growth rates, obtained at high constant K(max) to minimize crack closure and in vacuum to eliminate any environmental effect, were alloy dependent; da/dN varied up to tenfold based on applied Delta-K or Delta-K/E. When compared based on a crack tip cyclic strain or opening displacement parameter, growth rates were equivalent for all alloys except X2095-T8, which exhibited unique fatigue crack growth resistance. Tortuous fatigue crack profiles and large fracture surface facets were observed for each Al-Li alloy independent of the precipitates present, particularly delta-prime, and the localized slip deformation structure. Reduced fatigue crack propagation rates for X2095 in vacuum are not explained by either residual crack closure or slip reversibility arguments; the origin of apparent slip band facets in a homogeneous slip alloy is unclear.

  20. Effect of heat-treatment on fatigue property of Al-Li alloy

    Institute of Scientific and Technical Information of China (English)

    张荻; 丁剑; 范同祥; 吕维洁; 覃继宁

    2003-01-01

    Fatigue property of Al-Li alloy after various heat treatment was investigated. The results show that the fatigue strength is enhanced with the age hardening progressing. Compared to the solution treated specimen, the fatigue limit is improved to 136% for sub-ageing treated specimen and 155% for peak-ageing treated specimen, respectively. In the meanwhile, the fatigue deformation becomes non-uniform with age hardening progressing. The fatigue cracks initiate and propagate prior from the un-uniform slip band, causing transgranular fracture or the mixed mode of transgranular fracture and intergranular fracture.

  1. Influence of Electric Field on Mechanical Properties of Al-Li Alloy Containing Cerium and Electronic Mechanism

    Institute of Scientific and Technical Information of China (English)

    刘兵; 陈铮; 王永欣; 王西宁

    2001-01-01

    The effect of electric field on the mechanical properties and microstructure of Al-Li alloy containing Ce was investigated, and mechanism was discussed. The experimental results show that the ductility of the alloy is enhanced by the electric field. The fracture features are changed and the precipitates are dispersed under the effect of the electric field. The mechanism discussion reveals that the effects of the electric field on the alloy are due to the change of the electron density in the alloy.

  2. Effects of Annealing Process on the Formability of Friction Stir Welded Al-Li Alloy 2195 Plates

    Science.gov (United States)

    Chen, Po-Shou; Bradford, Vann; Russell, Carolyn

    2011-01-01

    Large rocket cryogenic tank domes have typically been fabricated using Al-Cu based alloys like Al-Cu alloy 2219. The use of aluminum-lithium based alloys for rocket fuel tank domes can reduce weight because aluminum-lithium alloys have lower density and higher strength than Al-Cu alloy 2219. However, Al-Li alloys have rarely been used to fabricate rocket fuel tank domes because of the inherent low formability characteristic that make them susceptible to cracking during the forming operations. The ability to form metal by stretch forming or spin forming without excessive thinning or necking depends on the strain hardening exponent "n". The stain hardening exponent is a measure of how rapidly a metal becomes stronger and harder. A high strain hardening exponent is beneficial to a material's ability to uniformly distribute the imposed strain. Marshall Space Flight Center has developed a novel annealing process that can achieve a work hardening exponent on the order of 0.27 to 0.29, which is approximately 50% higher than what is typically obtained for Al-Li alloys using the conventional method. The strain hardening exponent of the Al-Li alloy plates or blanks heat treated using the conventional method is typically on the order of 0.17 to 0.19. The effects of this novel annealing process on the formability of friction stir welded Al-Li alloy blanks are being studied at Marshall Space Flight Center. The formability ratings will be generated using the strain hardening exponent, strain rate sensitivity and forming range. The effects of forming temperature on the formability will also be studied. The objective of this work is to study the deformation behavior of the friction stir welded Al-Li alloy 2195 blank and determine the formability enhancement by the new annealing process.

  3. Simultaneous SAS and 100 experiments on phase decomposition and reversion in Al-Li binary alloys

    International Nuclear Information System (INIS)

    Phase decomposition and reversion processes in Al-Li binary alloys have been studied by synchrotron-radiation small-angle/100 scattering experiments. The microstructure and its evolution obtained from small-angle scattering (SAS) and 100 profiles during phase decomposition and reversion are discussed. For the coarsening and reversion processes where a well defined interface between the δ' precipitates and the matrix can be expected, the information obtained from the SAS and 100 profiles was essentially the same. On the other hand, the structural information they convey can be different in the early stage of phase decomposition. The interpretation of the SAS and 100 intensities by means of an extension of the two-phase model has been examined. (orig.)

  4. Strength and microstructure of 2091 Al-Li alloy TIG welded joint

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The microstructure and tensile properties of TIG welding joints of 2091 Al-Li alloy were investigated both in as-welded and different postweld heat treatment condition. The results show that solution strengthening played an important role in the as-welded condition, though the precipitation strengthening δ' phase formed already in the as-welded weld metal, but its effect was not apparent due to the lower volume fraction of δ' phase. So the strength coefficient (φ) of the welded joint/base metal was 64%. After artificially aging heat treatment, the precipitation strengthening effect increased much due to the formation of more δ' phase and s' phase. Its φ value was increased up to 89%. The highest strength of the welded joints was obtained after solid solution and then artificially aged heat treatment. Due to the proper size of precipitation strengthening phases and their well distribution, the φ value was increased up to 98%.

  5. The influence of heat treatment on the properties of laser welded Al-Li alloy

    International Nuclear Information System (INIS)

    An Al-Li-Cu-Mg-Zr alloy was welded with CO2 laser to produce full-penetration, single pass butt welds. Initial YS of 487 MPa for unwelded material decreased to 268 MPa after laser welding as well as the measured elongation decreased from 17.48% to 3.7% respectively. The postweld heat treatment consisting of solutionizing at 550 oC for 2 h and/or artificial aging at 150, 175, 200 and 225oC for 2, 4, 8, 16 and 32 h were performed to improve the properties of welded material. Studies by optical, scanning and transmission electron microscopes were provided in: as-welded, as-heat-treated and as-deformed states to show the microstructural changes with postweld heat treatment. (author)

  6. Evolution of grain structure in AA2195 Al-Li alloy plate during recrystallization

    Institute of Scientific and Technical Information of China (English)

    DU Yu-xuan; ZHANG Xin-ming; YE Ling-ying; LIU Sheng-dan

    2006-01-01

    The evolution of the grain structures in AA2195 Al-Li alloy plate warm-rolled by 80% reduction during recrystallization annealing at 500 ℃ was investigated by electron backscatter diffraction, scanning electron microscopy and transmission electron microscopy. It is found that the elongated grain structures are caused by the lamellar distribution of recrystallization nucleation sites,being lack of large second phase particles (> 1 μm), and dispersive coherent particles (such as δ'andβ) concentrated in planar bands.The recrystallization process may be separated into three stages: firstly, recrystallization nucleation occurs heterogeneously, and the nuclei are concentrated in some planar zones parallel to rolling plane. Secondly, the grain boundaries interacted with small particles concentrate in planar bands, which is able to result in the elongated grain structures. The rate of the grain growth is controlled by the dissolution of these small particles. Thirdly, after most of small particles are dissolved, their hindrance to migration of the grain boundaries fades away, and the unrecrystallized zones are consumed by adjacent recrystallized grains. The migration of high angle grain boundaries along normal direction leads a gradual transformation from the elongated grains to the nearly equiaxed, which is driven by the tension of the grain boundaries.

  7. A binary Al/Li alloy as a new material for the realization of high-intensity pulsed photocathodes

    Science.gov (United States)

    Septier, A.; Sabary, F.; Dudek, J. C.; Bergeret, H.; Leblond, B.

    1991-07-01

    We propose a new material for the fabrication of high-current photocathodes: a binary Al/Li alloy acting as a lithium dispenser cathode. This material would have the great advantage to allow regeneration of the Li layer after poisoning or air exposure, by a simple heating process. In a first experiment, we have measured the photoemission energy threshold, WΦ, of a piece of Al/Li alloy and the quantum yield, Y, as a function of the photon energy. After a heating process (340°C for 12 h) we obtained WΦ = 2 eV and Y = 6 × 10 -4 for 4.6 eV photon energy. In a second experiment another sample was illuminated with a 40 ps frequency-tripled YAG laser. After two heating processes, we obtained electron bunches containing 1 nC with an incident laser energy of 100 μJ per pulse.

  8. Process Optimization of Dual-Laser Beam Welding of Advanced Al-Li Alloys Through Hot Cracking Susceptibility Modeling

    Science.gov (United States)

    Tian, Yingtao; Robson, Joseph D.; Riekehr, Stefan; Kashaev, Nikolai; Wang, Li; Lowe, Tristan; Karanika, Alexandra

    2016-07-01

    Laser welding of advanced Al-Li alloys has been developed to meet the increasing demand for light-weight and high-strength aerospace structures. However, welding of high-strength Al-Li alloys can be problematic due to the tendency for hot cracking. Finding suitable welding parameters and filler material for this combination currently requires extensive and costly trial and error experimentation. The present work describes a novel coupled model to predict hot crack susceptibility (HCS) in Al-Li welds. Such a model can be used to shortcut the weld development process. The coupled model combines finite element process simulation with a two-level HCS model. The finite element process model predicts thermal field data for the subsequent HCS hot cracking prediction. The model can be used to predict the influences of filler wire composition and welding parameters on HCS. The modeling results have been validated by comparing predictions with results from fully instrumented laser welds performed under a range of process parameters and analyzed using high-resolution X-ray tomography to identify weld defects. It is shown that the model is capable of accurately predicting the thermal field around the weld and the trend of HCS as a function of process parameters.

  9. Aluminum alloys for ALS cryogenic tanks: Comparative measurements of cryogenic mechanical properties of Al-Li alloys and alloy 2219, February 1993

    International Nuclear Information System (INIS)

    Tensile and fracture toughness were obtained at cryogenic temperatures to compare the Al-Li alloys 8090, 2090, and WL049, and alloy 2219 in various tempers and specimen orientations. The strongest alloy at very low temperatures is WL049-T851, which is about 10 percent stronger than 2090-T81. Both alloys are considerably stronger than 2219-T87. Alloy 2090-T81 is tougher (about 50 percent) than WL049-T851 at low temperatures; the higher toughness is attributed to the presence of fewer constituent particles and the tendency to delaminate at low temperatures. The delamination divides the moving crack, thus separating it into smaller regions where plane stress (rather than plane strain) conditions are conducive to increased toughness

  10. Effect of specimen orientation and welding on the fracture and fatigue properties of 2195 Al-Li alloy

    International Nuclear Information System (INIS)

    In view of the use of 2195 Al-Li alloy in the construction of super-light-weight external fuel tank of space shuttles, bulkheads of reusable single-stage-to-orbit launch vehicles and in combat ground vehicles, the dependence of tensile properties, fracture toughness and fatigue resistance of this alloy on the specimen orientation and welding is very important and was studied. The T8 base alloy, with primary strengthening precipitates of T1 (Al2CuLi) phase, contained mainly brass-type texture. After welding with AA 4043 filler alloy, the fusion zone (FZ) consisted of T (AlLiSi) phase and in the heat-affected zone (HAZ) T1 phase was replaced by TB (Al7Cu4Li) phase, and micro-cracks were observed. The post-weld heat treatment (PWHT) resulted in the spheroidization of primary T phase and the precipitation of more T particles in the FZ, and the dissolution of TB phase and the re-precipitation of T1 phase in the HAZ. The yield strength, fracture toughness and fatigue threshold of the 2195-T8 alloy was observed to depend on the specimen orientation, with the lowest values obtained at 45 deg. to the rolling direction. Welding resulted in a reduction in the tensile properties and fatigue strength. The post-weld heat treatment enhanced the yield strength, but no increase in fatigue strength was observed. Fracture mechanisms in various cases were evaluated by SEM examination of fracture surfaces and are discussed

  11. Study on electrochemical preparation of Al-Li-Y alloys from Y2O3 in LiCl-KCl-AlCl3 molten salts

    Institute of Scientific and Technical Information of China (English)

    LI Yaming; WANG Fengli; ZHANG Milin; HAN Wei; TIAN Yang

    2011-01-01

    The electrochemical preparaton of Al-Li-Y alloys from LiCl-KCl-A1Cl3-Y2O3 system was studied. The chlorination of Y2O3 by AlCl3 led to the formation of Y (Ⅲ) ions in the molten salts. Cyclic voltammogram (CV) showed that the underpotential deposition (UPD) of yttrium on pre-deposited aluminum caused the formation of Al-Y alloy. Al-Li-Y alloys with different yttriurn contents were obtained by galvanostatic electrolysis and analysed by SEM-EDS and ICP. The ICP results showed that the lithium and yttrium contents in Al-Li-Y alloysdepended on the addition of AlCl3 into the melts.

  12. The effect of zinc additions on the environmental stability of Alloy 8090 (Al-Li-Cu-Mg-Zr)

    Science.gov (United States)

    Kilmer, Raymond J.; Stoner, G. E.

    1991-01-01

    Stress corrosion cracking (SCC) remains a problem in both Al-Li and conventional Al heat treatable alloys. It has recently been found that relatively small additions (less than or approximately 1 wt-percent) of Zn can dramatically improve the SCC performance of alloy 8090 (Al-Li-Cu-Mg-Zr). Constant load time to failure experiments using cylindrical tensile samples loaded between 30 and 85 percent of TYS indicate improvements of orders of magnitude over the baseline 8090 for the Zn-containing alloys under certain aging conditions. However, the toughnesses of the alloys were noticeably degraded due to the formation of second phase particles which primarily reside on grain and subgrain boundaries. EDS revealed that these intermetallic particles were Cu and Zn rich. The particles were present in the T3 condition and were not found to be the result of quench rate, though their size and distribution were. At 5 hours at 160 C, the alloys displayed the greatest susceptibility to SCC but by 20 hours at 160 C the alloys demonstrated markedly improved TTF lifetimes. Aging past this time did not provide separable TTF results, however, the alloy toughnesses continued to worsen. Initial examination of the alloys microstructures at 5 and 20 hours indicated some changes most notably the S' and delta' distributions. A possible model by which this may occur will be explored. Polarization experiments indicated a change in the trend of E(sub BR) and passive current density at peak aging as compared to the baseline 8090. Initial pitting experiments indicated that the primary pitting mechanism in chloride environments is one occurring at constituent (Al-Fe-Cu) particles and that the Cu and Zn rich boundary precipitates posses a breakaway potential similar to that of the matrix acting neither anodic or cathodic in the first set of aerated 3.5 w/o NaCl experiments. Future work will focus on the identification of the second phase particles, evaluation of K(sub 1SCC) and plateau da/dt via

  13. Effect of electric current pulse on grain growth in superplastic deformation of 2091 Al-Li alloy

    Institute of Scientific and Technical Information of China (English)

    刘志义; 许晓嫦; 崔建忠

    2003-01-01

    The effect of electric current pulse on the grain growth in the superplastic deformation of 2091 Al-Li alloy was investigated. Optical metallographic microstructure observation and average linear intercept measuring results show that at same strain, the grain size in the superplastic deformation loaded with electric current pulse is smaller than that unemploying electric current pulse, and so does the grain growth rate. TEM observation shows that the dislocation density at grain boundary in the superplastic deformation applied with electric current pulse is lower than that unemploying electric current pulse.It indicates that electric current pulse increases the rate of dislocation slip and climb in grain boundary, which leads to a decrease of both the density of the dislocation slipping across grain boundary at same strain rate and the driving force for grain growth, therefore the rate of grain growth decreases.The established model for grain growth shows an exponential relation of grain size with strain.

  14. Environmental fatigue of an Al-Li-Cu alloy. Part 3: Modeling of crack tip hydrogen damage

    Science.gov (United States)

    Piascik, Robert S.; Gangloff, Richard P.

    1992-01-01

    Environmental fatigue crack propagation rates and microscopic damage modes in Al-Li-Cu alloy 2090 (Parts 1 and 2) are described by a crack tip process zone model based on hydrogen embrittlement. Da/dN sub ENV equates to discontinuous crack advance over a distance, delta a, determined by dislocation transport of dissolved hydrogen at plastic strains above a critical value; and to the number of load cycles, delta N, required to hydrogenate process zone trap sites that fracture according to a local hydrogen concentration-tensile stress criterion. Transgranular (100) cracking occurs for process zones smaller than the subgrain size, and due to lattice decohesion or hydride formation. Intersubgranular cracking dominates when the process zone encompasses one or more subgrains so that dislocation transport provides hydrogen to strong boundary trapping sites. Multi-sloped log da/dN-log delta K behavior is produced by process zone plastic strain-hydrogen-microstructure interactions, and is determined by the DK dependent rates and proportions of each parallel cracking mode. Absolute values of the exponents and the preexponential coefficients are not predictable; however, fractographic measurements theta sub i coupled with fatigue crack propagation data for alloy 2090 established that the process zone model correctly describes fatigue crack propagation kinetics. Crack surface films hinder hydrogen uptake and reduce da/dN and alter the proportions of each fatigue crack propagation mode.

  15. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  16. Environment and microstructure effects on fatigue crack facet orientation in an Al-Li-Cu-Zr alloy

    International Nuclear Information System (INIS)

    The effects of environment, microstructure and texture on transgranular fatigue crack facet orientation are established with electron-back scattered pattern analysis and stereofractography for single grains in peak aged Al-Li-Cu-Zr alloy 2090. For vacuum, facets are near-{111} due to fatigue fracture through intense deformation bands with a complex planar-slip dislocation structure. Multiple facets in single grains and the tortuous crack path are caused by high shear stresses resolved on multiple slip systems. Low stress intensity range fatigue fracture in NaCl is transgranular and faceted, but not tortuous. Eighty-five percent of the facets in unrecrystallized plate and 50% of the facets in recrystallized sheet are within 10 of a high index plane, on average {521}, subjected to high normal stresses. Such facets are inconsistent with: (a) hydrogen-enhanced localized plasticity and {111} decohesion; (b) slip-locking with bisecting {100} cracking; (c) environment-enhanced alternate slip with {100} faceting; or (d) {100}/{100} decohesion. Environmental fatigue may be governed by faceted cracking associated with hydrides or hydrogen embrittled dislocation cell walls

  17. THE EFFECT OF HYDROGEN DURING STRESS CORROSION CRACKING AND CORROSION FATIGUE OF Al-Li-Cu ALLOYS IN 3.5 % NaCl SOLUTIONS

    OpenAIRE

    Magnin, T; RebiÈre, M.

    1987-01-01

    Stress corrosion cracking and corrosion fatigue tests at imposed strain rate are conducted on an industrial Al-Li-Cu alloy in a 3.5 % NaCl solution at imposed potential. In the condition of the tests and for short fatigue lifetimes, the anodic dissolution is shown to play the predominant role during corrosion-fatigue at free corrosion potential. Nevertheless a marked hydrogen embrittlement is observed at cathodic potentials during corrosion fatigue of specimens containing superficial microcra...

  18. Environmental fatigue of an Al-Li-Cu alloy. Part 1: Intrinsic crack propagation kinetics in hydrogenous environments

    Science.gov (United States)

    Piascik, Robert S.; Gangloff, Richard P.

    1991-01-01

    Deleterious environmental effects on steady-state, intrinsic fatigue crack propagation (FCP) rates (da/dN) in peak aged Al-Li-Cu alloy 2090 are established by electrical potential monitoring of short cracks with programmed constant delta K and K(sub max) loading. The da/dN are equally unaffected by vacuum, purified helium, and oxygen but are accelerated in order of decreasing effectiveness by aqueous 1 percent NaCl with anodic polarization, pure water vapor, moist air, and NaCl with cathodic polarization. While da/dN depends on delta K(sup 4.0) for the inert gases, water vapor and chloride induced multiple power-laws, and a transition growth rate 'plateau'. Environmental effects are strongest at low delta K. Crack tip damage is ascribed to hydrogen embrittlement because of the following: (1) accelerated da/dN due to part-per-million levels of H2O without condensation; (2) impeded molecular flow model predictions of the measured water vapor pressure dependence of da/dN as affected by mean crack opening; (3) the lack of an effect of film-forming O2; (4) the likelihood for crack tip hydrogen production in NaCl, and (5) the environmental and delta K-process zone volume dependencies of the microscopic cracking modes. For NaCl, growth rates decrease with decreasing loading frequency, with the addition of passivating Li2CO3, and upon cathodic polarization. These variables increase crack surface film stability to reduce hydrogen entry efficiency. The hydrogen environmental FCP resistance of 2090 is similar to other 2000 series alloys and is better than 7075.

  19. Environmental fatigue of an Al-Li-Cu alloy. I - Intrinsic crack propagation kinetics in hydrogenous environments

    Science.gov (United States)

    Piascik, Robert S.; Gangloff, Richard P.

    1991-01-01

    Deleterious environmental effects on steady-state, intrinsic fatigue crack propagation (FCP) rates (da/dN) in peak aged Al-Li-Cu alloy 2090 are established by electrical potential monitoring of short cracks with programmed constant delta K and K(sub max) loading. The da/dN are equally unaffected by vacuum, purified helium, and oxygen but are accelerated in order of decreasing effectiveness of aqueous 1 percent NaCl with anodic polarization, pure water vapor, moist air, and NaCl with cathodic polarization. While da/dN depends on delta K(sup 4.0) for the inert gases, water vapor and chloride induced multiple power-laws, and a transition growth rate 'plateau'. Environmental effects are strongest at low delta K. Crack tip damage is ascribed to hydrogen embrittlement because of the following: (1) accelerated da/dN due to part-per-million levels of H2O without condensation; (2) impeded molecular flow model predictions of the measured water vapor pressure dependence of da/dN as affected by mean crack opening; (3) the lack of an effect of film-forming O2; (4) the likelihood for crack tip hydrogen production in NaCl; and (5) the environmental and delta K-process zone volume dependencies of the microscopic cracking modes. For NaCl, growth rates decrease with decreasing loading frequency, with the addition of passivating Li2CO3, and upon cathodic polarization. These variables increase crack surface film stability to reduce hydrogen entry efficiency. The hydrogen environmental FCP resistance of 2090 is similar to other 2000 series alloys and is better than 7075.

  20. LASER CLADDING ON ALUMINIUM BASE ALLOYS

    OpenAIRE

    Pilloz, M.; Pelletier, J; Vannes, A.; Bignonnet, A.

    1991-01-01

    laser cladding is often performed on iron or titanium base alloys. In the present work, this method is employed on aluminum alloys ; nickel or silicon are added by powder injection. Addition of silicon leads to sound surface layers, but with moderated properties, while the presence of nickel induces the formation of hard intermetallic compounds and then to an attractive hardening phenomena ; however a recovery treatment has to be carried out, in order to eliminate porosity in the near surface...

  1. Laser-induced reversion of $\\delta^{'}$ precipitates in an Al-Li alloy: Study on temperature rise in pulsed laser atom probe

    CERN Document Server

    Khushaim, Muna; Al-Kassab, Talaat

    2015-01-01

    The influence of tuning the laser energy during the analyses on the resulting microstructure in a specimen utilizing an ultra-fast laser assisted atom probe was demonstrated by a case study of a binary Al-Li alloy. The decomposition parameters, such as the size, number density, volume fraction and composition of $\\delta^{'}$ precipitates, were carefully monitored after each analysis. A simple model was employed to estimate the corresponding specimen temperature for each value of the laser energy. The results indicated that the corresponding temperatures for the laser energy in the range of 10 to 80 pJ are located inside the miscibility gap of the binary Al-Li phase diagram and fall into the metastable equilibrium field. In addition, the corresponding temperature for a laser energy of 100 pJ was in fairly good agreement with reported range of $\\delta^{'}$ solvus temperature, suggesting a result of reversion upon heating due to laser pulsing.

  2. Evaluation of Engineering Properties of AL-Li Alloy X2096-T8A3 Extrusion Products

    Science.gov (United States)

    Flom, Y.; Viens, M.; Wang, L.

    1999-01-01

    Mechanical, thermal fatigue and stress corrosion properties were determined for the two lots of Al-Li X2096-T8A3 extruded beams. Based on the test results, the beams were accepted as the construction material for fabrication of the Hubble Space Telescope new Solar Array Support Structure.

  3. Structure and properties of a rapidly solidified Al-Li-Mn-Zr Alloy for high-temperature applications: Part I. inert gas atomization processing

    Science.gov (United States)

    Ruhr, Michael; Baram, Joseph

    1991-10-01

    A new Al-Li alloy containing 2.3 wt pct Li, 6.5 wt pct Mn, and 0.65 wt pet Zr, for high-temperature applications, has been processed by a rapid solidification (RS) technique (as powders by inert gas atomization) and then thermomechanically treated by hot isostatic pressing (hipping) and hot extrusion. As-received and thermomechanically treated powders (of various size fractions) were characterized by X-ray diffraction and scanning and transmission electron microscopy (SEM and TEM, respectively). Phase analyses in the as-processed materials revealed the presence of two Mn phases (Al4Mn and Al6Mn), one Zr phase (Al3Zr), two Li phases (the stable AlLi and the metastable Al3Li), and the αAl solid solution with high excess in Mn solubility (up to close the nominal composition in the as-atomized powders). Extruded pieces were solutionized at 370 °C and 530 °C for various soaking times (2 to 24 hours). A variety of aging treatments was practiced to check for the optimal (for tensile properties) aging procedure, which was found to be the following: solutioning at 370 °C for 2 hours and water quenching + 1 pct mechanical stretching + one step aging at 120 °C for 3 hours. The mechanical properties, at room and elevated temperatures, of the “hipped” and hot extruded powders are compared following the optimal solutioning and aging treatments. The results indicate that Mn is indeed a favorable alloying element for rapidly solidified Al-Li alloys to retain about 85 to 95 pct of the room-temperature tensile properties even at 250 °C, though room-temperature strength is not satisfactory in itself. However, specific moduli are by 20 to 25 pet higher than those of the 2024 series duralumin-type alloys. Ductilities at room temperatures are in the low 1 to 2.5 pct range and show no improvement over other Al-Li alloys.

  4. Structure and properties of a rapidly solidified Al-Li-Mn-Zr alloy for high-temperature applications: Part II. spray atomization and deposition processing

    Science.gov (United States)

    Baram, Joseph

    1991-10-01

    A new Al-Li alloy containing 2.3 wt pct Li, 6.5 wt pct Mn, and 0.65 wt pet Zr for high-temperature applications has been processed by a rapid solidification (RS) technique (as compacts by spray atomization and deposition) and then thermomechanically treated by hot extrusion. As-received and thermomechanically treated deposits were characterized by X-ray diffraction and scanning electron microscopy (SEM). Phase analyses in the as-processed materials revealed the presence of two Mn phases (Al4Mn and Al6Mn), one Zr phase (Al3Zr), two Li phases (the stable AlLi and the metastable Al3Li), and the aAl solid solution with high excess in Mn solubility (up to close the nominal composition in the as-atomized powders). As-deposited and extruded pieces were given heating treatments at 430 °C and 530 °C. A two-step aging treatment was practiced, to check for the optimal (for tensile properties) aging procedure, which was found to be the following: solutioning at 430 °C for 1 hour and water quenching + a first-step aging at 120 °C for 12 hours + a second-step aging at 175 °C for 15 hours. The mechanical properties, at room and elevated temperatures, of the hot extruded deposits are compared, following the optimal solutioning and aging treatments. The room-temperature (RT) strength of the proposed alloy is distinctly better for the as-deposited specimens (highest yield strength, 320 MPa) than for the as-atomized (highest yield strength, 215 MPa), though less than 65 pct of the RT strength is conserved at 250 °C. Ultimate strengths are quite comparable (in the 420 to 470 MPa range). Ductilities at RTs are in the low 1.5 to 2.5 pct range and show no improvement over other Al-Li alloys.

  5. Behavior of Growth and Coarsening of T1(Al2CuLi) Precipitates in Al-Li Alloys 2090 and 2090+Ce

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The behavior of growth and coarsening of T1(Al2CuLi) precipitates was comparatively studied by means of TEM technique in two Al-Li alloys 2090 and 2090+Ce (with cerium content less than 0.1% in mass fraction). Statistical analysis results show that T1 precipitates in alloy 2090+Ce have smaller aspect ratio, which is connected with the more intersections between different T1 variants in this alloy. It is also found that the variation of maximum length of T1 precipitates with aging time can be obviously divided into two stages of growth and coarsening. The diffusion coefficients of solute atoms of Cu and Li are calculated via growth kinetics curves of T1 precipitates. The results show that the diffusion of atom Cu plays a more important role in the formation of T1 precipitates.

  6. Investigation of Abnormal Grain Growth in a Friction Stir Welded and Spin-Formed Al-Li Alloy 2195 Crew Module

    Science.gov (United States)

    Tayon, Wesley A.; Domack, Marcia S.; Hoffman, Eric K.; Hales, Stephen J.

    2013-01-01

    In order to improve manufacturing efficiency and reduce structural mass and costs in the production of launch vehicle structures, NASA is pursuing a wide-range of innovative, near-net shape manufacturing technologies. A technology that combines friction stir welding (FSW) and spin-forming has been applied to manufacture a single-piece crew module using Aluminum-Lithium (AL-Li) Alloy 2195. Plate size limitations for Al-Li alloy 2195 require that two plates be FSW together to produce a spin-forming blank of sufficient size to form the crew module. Subsequent forming of the FSW results in abnormal grain growth (AGG) within the weld region upon solution heat treatment (SHT), which detrimentally impacts strength, ductility, and fracture toughness. The current study seeks to identify microstructural factors that contribute to the development of AGG. Electron backscatter diffraction (EBSD) was used to correlate driving forces for AGG, such as stored energy, texture, and grain size distributions, with the propensity for AGG. Additionally, developmental annealing treatments prior to SHT are examined to reduce or eliminate the occurrence of AGG by promoting continuous, or uniform, grain growth

  7. Ferrous Alloy Powder for Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    WEN Jialing; NIU Quanfeng; XU Yanmin

    2005-01-01

    This investigation aimed at how to improve the hardness and wear resistance by B, Si and Cr, and how to improve the synthesis property by Re (rare-earth element). Based on the experiment of Fe-based alloys of Fe-Cr-Ni-B-Si-Re, through experiments and a serious of synthesis analysis, including surface quality, spectrum composite, micro-hardness, scanning electron microscopy, as well as the synthesis evaluation,etc, prescriptions were optimized. As a result, a Fe-Cr-Ni-B-Si-Re cladding material with a high property was obtained.

  8. LASER CLADDING WITH COBALT-BASED HARDFACING ALLOYS

    OpenAIRE

    Frenk, A.; WagniÈre, J.-D.

    1991-01-01

    Preliminary results aimed at designing Co-based hardfacing alloys specifically for the laser cladding process are reported. Three alloys, ranging from hypo- to hypereutectic were deposited using scanning velocities between 1.7 and 170 mm/s. The microstructures and the dry sliding wear resistances of the clads were investigated. First trends relating composition to dry sliding wear resistance were deduced.

  9. Effect of T6I6 and its modified processes on mechanical properties of novel high strength Al-Li alloy-2A97

    Institute of Scientific and Technical Information of China (English)

    YUAN Zhi-shan; LU Zheng; XIE You-hua; WU Xiu-liang; DAI Sheng-long; LIU Chang-sheng

    2006-01-01

    Based on a novel high strength Al-Cu-Li-X alloy-2A97, the effect of T6I6 and its modified processes on the properties investigated by SEM and tensile test. The results show that when the alloy is heat treated by triple ageing, with secondary low temperature ageing at 80 ℃ after initial ageing at 155 and 150 ℃, and final re-ageing at 135 and 165 ℃, the tensile properties are close to the peak level of aged alloy in T6 temper. The addition of plastic deformation after and prior to secondary ageing favor the T1(Al2CuLi) and δ'(Al3Li) precipitation during final re-ageing at 135 and 165 ℃ corresponding to triple ageing, so the Al-Li alloy displays higher strength for the modified processes of T6I6. The microstructures consist of δ', T1 and θ"/θ' (Al2Cu) phase for single and triple aged alloy, the number density and volume fraction of δ' phase increase for T6I6 and its modified processes correspond to single ageing.

  10. Laser cladding of titanium alloy coating on titanium aluminide alloy substrate

    Institute of Scientific and Technical Information of China (English)

    徐子文; 黄正; 阮中健

    2003-01-01

    A new diffusion bonding technique combined with laser cladding process was developed to join TiAl alloy to itself and Ti-alloys. In order to enhance the weldability of TiAl alloys, Ti-alloy coatings were fabricated by laser cladding on the TiAl alloy. Ti powder and shaped Ti-alloy were respectively used as laser cladding materials. The materials characterization was carried out by OM, SEM, EDS and XRD analysis. The results show that the laser cladding process with shaped Ti-alloy remedy the problems present in the conventional process with powder, such as impurities, cracks and pores. The diffusion bonding of TiAl alloy with Ti-alloy coating to itself and Ti-alloy was carried out with a Gleeble 1500 thermal simulator. The sound bonds of TiAl/TiAl, TiAl/Ti were obtained at a lower temperature and with shorter time.

  11. Influence of aging at 180C on the corrosion behaviour of a ternary Al-Li-Zr alloy

    DEFF Research Database (Denmark)

    Ambat, Rajan; Prasad, R.K.; Dwarakadasa, E.S.

    1994-01-01

    The influence of aging at 180 °C on the corrosion behaviour of an Al-1.5%Li-0.1%Zr alloy has been studied using weight loss, open circuit potential (OCP) measurements and potentiodynamic polarization measurements in 3.5% NaCl solution. Corrosion rates obtained from weight loss and Icorr values...... current density showed an increase. The results have been interpreted in terms of the microstructure....

  12. Preparation and Cycling Performance of Iron or Iron Oxide Containing Amorphous Al-Li Alloys as Electrodes

    Directory of Open Access Journals (Sweden)

    Franziska Thoss

    2014-12-01

    Full Text Available Crystalline phase transitions cause volume changes, which entails a fast destroying of the electrode. Non-crystalline states may avoid this circumstance. Herein we present structural and electrochemical investigations of pre-lithiated, amorphous Al39Li43Fe13Si5-powders, to be used as electrode material for Li-ion batteries. Powders of master alloys with the compositions Al39Li43Fe13Si5 and Al39Li43Fe13Si5 + 5 mass-% FeO were prepared via ball milling and achieved amorphous/nanocrystalline states after 56 and 21.6 h, respectively. In contrast to their Li-free amorphous pendant Al78Fe13Si9, both powders showed specific capacities of about 400 and 700 Ah/kgAl, respectively, after the third cycle.

  13. Phase separation and ordering process in Al-Li alloys studied by small-angle neutron scattering and neutron diffraction

    International Nuclear Information System (INIS)

    To study phase separation kinetics of Al-9.5at.%Li polycrystalline alloys in which precipitates have ordered Al3Li (δ') structure, profile analysis of small-angle neutron scattering and superlattice reflections (100) and (110) were done. A small-angle scattering instrument and a triple-axis spectrometer in elastic mode were used in the measurements. Strong texture was observed in the reflections. Therefore, measurements were done using the crystal orientation where the intensity of the reflection was at the maximum. Profiles of small-angle scattering and superlattice reflections were almost identical at higher momentum transfer side. At lower momentum transfer side, small-angle scattering showed interference effects, but superlattice reflection did not show any sign of interference. Integrated intensities of superlattice reflections were obtained and compared with small-angle scattering intensity. The order parameter was not saturated in the δ' precipitates at the early stage of the phase separation process

  14. Calculation of electric field effects on the Gibbs free energy of the Al-Li-Mg alloy

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Based on the Thomas-Fermi model the calculation methods of the monatomic Gibbs free energy were found.The interior potential boundary condition under electric field was defined. The calculation methods of Gibbs free energy for the monatomic, compound, and solid solution with electric field were set up. Under the influence of electric field, the Gibbs free energy of A1 is the most sensitive, followed by those of Li and Mg. At the solution temperature the Gibbs free energies of Al3Li and its elements under electric field are not symmetrical about the zero point of electric field, whereas at the aging temperature their values are symmetrical about the zero point of electric field. At the solution temperature near the zero point of electric field, the Gibbs free energy of Al3Li is higher than that of Al-2.14%Li. And at 460 K the Gibbs free energy of A13Li is lower than that of Al-2.14wt.%Li under electric field. The Gibbs free energy of 1420 alloy decreases from both sides of electric field to the zero point at the aging temperature.

  15. Characterization of Precipitation in Al-Li Alloy AA2195 by means of Atom Probe Tomography and Transmission Electron Microscopy

    KAUST Repository

    Khushaim, Muna

    2015-05-19

    The microstructure of the commercial alloy AA2195 was investigated on the nanoscale after conducting T8 tempering. This particular thermomechanical treatment of the specimen resulted in the formation of platelet-shaped T 1 Al 2 CuLi / θ ′ Al 2 Cu precipitates within the Al matrix. The electrochemically prepared samples were analyzed by scanning transmission electron microscopy and atom probe tomography for chemical mapping. The θ ′ platelets, which are less than 2 nm thick, have the stoichiometric composition consistent with the expected Al 2 Cu equilibrium composition. Additionally, the Li distribution inside the θ ′ platelets was found to equal the same value as in the matrix. The equally thin T 1 platelet deviates from the formula (Al 2 CuLi) in its stoichiometry and shows Mg enrichment inside the platelet without any indication of a higher segregation level at the precipitate/matrix interface. The deviation from the (Al 2 CuLi) stoichiometry cannot be simply interpreted as a consequence of artifacts when measuring the Cu and Li concentrations inside the T 1 platelet. The results show rather a strong hint for a true lower Li and Cu contents, hence supporting reasonably the hypothesis that the real chemical composition for the thin T 1 platelet in the T8 tempering condition differs from the equilibrium composition of the thermodynamic stable bulk phase.

  16. Coarsening kinetics, thermodynamic properties, and interfacial characteristics of δ' precipitates in Al-Li alloys taking into account the Gibbs-Thomson effect

    International Nuclear Information System (INIS)

    The structure factor model of small-angle x-ray scattering (SAXS) analysis is validated herein by transmission electron microscopy (TEM) result regarding the volume fraction and size of δ' precipitates. The kinetic behaviors of the number density and volume fraction of δ' precipitates in Al-Li alloys during the coarsening stage are quantitatively investigated by SAXS. The results indicate that the conventional kinetic law must be replaced by a more general equation that incorporates the Gibbs-Thomson effect and the time-dependence of the volume fraction during Ostwald ripening. This work also proposes new methods that combine the Gibbs-Thomson effect and the traditional SAXS equation to resolve more reliably and model independently the interfacial energy, the concentration of solute Li in the α matrix in equilibrium with δ' particles of a nanoscale radius Cαr, the equilibrium solubility of the α phase Ceα and the equilibrium concentration of δ' particles. The Gibbs-Thomson effect considers the effects of the interfacial energy and particle size on the equilibrium concentration. This effect quantitatively clarifies that the Cαr value is size-dependent and is related to the Ceα value and the interfacial energy. The traditional SAXS equation determines the Li concentrations in the δ' particles and the matrix from the measured scattering contrast. The traditionally determined solubility is in fact the Cαr value and is mistakenly regarded as the equilibrium concentration Ceα (corresponding to the radius is infinite). These results are compared to other results obtained by SAXS, TEM, and calculation. The time evolution of the transition interfacial layers between δ' particles and the matrix is extensively investigated using SAXS

  17. Cladding Alloys for Fluoride Salt Compatibility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-05-01

    This interim report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for coating large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for coating inaccessible surfaces such as the interior surfaces of heat exchangers. The final project report will feature an experimental evaluation of the performance of the two selected cladding techniques.

  18. Residual stress measurements in laser clad aircraft aluminium alloys

    International Nuclear Information System (INIS)

    Fatigue and corrosion damage of structural components threatens the safety and availability of civil and military aircrafts. There is no sign of relief from these threats as civil and military aircrafts worldwide are continuously being pushed further into and past their initial design fatigue lives in tight financial circumstances. Given fatigue and corrosion damage often initiates at the surface and sub-surface of the components, there has been extensive research and development worldwide focused on advanced aircraft repair technologies and surface enhancement methods. The Deep Surface Rolling (DSR) is one of advanced surface enhancement technologies that can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. For the development of cost-effective aircraft structural repair technologies such as laser cladding, in this study, aluminium alloy 7075-T651 specimens with simulated corrosion damage were repaired using laser cladding technology. The surface of the laser cladding region was then processed by DSR. The experimental results from subsequent fatigue testing of laser cladded baseline, DSR and post-heat treated laser cladded specimens discovered that the DSR process can significantly increase fatigue life in comparison with the ascladded baseline. The three dimensional residual stresses were measured by neutron diffraction and the results confirmed the beneficial compressive residual stresses at the cladding surface can be achieved in depth more than 1.0 mm.

  19. DEVELOPMENT OF LASER CLADDING WEAR-RESISTANT COATING ON TITANIUM ALLOYS

    OpenAIRE

    RUILIANG BAO; HUIJUN YU; CHUANZHONG CHEN; BIAO QI; LIJIAN ZHANG

    2006-01-01

    Laser cladding is an advanced surface modification technology with broad prospect in making wear-resistant coating on titanium alloys. In this paper, the influences of laser cladding processing parameters on the quality of coating are generalized as well as the selection of cladding materials on titanium alloys. The microstructure characteristics and strengthening mechanism of coating are also analyzed. In addition, the problems and precaution measures in the laser cladding are pointed out.

  20. Corrosion properties of cladding materials from Zr1Nb alloy

    International Nuclear Information System (INIS)

    The corrosion behaviour was observed of the Zr1Nb alloy in hot water and superheated steam and the effects of impurity content, of the purity of the corrosion environment and of the heat treatment of the alloy were studied on the alloy corrosion resistance. Also studied were the absorption of hydrogen by the alloy and its behaviour in reactor situations. It was ascertained that the alloy has a good corrosion resistance up to a temperature of 350 degC. The corrosion resistance is reduced by the presence of nitrogen above 50 to 70 ppm and of carbon above 50 to 90 ppm. A graphic representation is given of the dependence of corrosion resistance on the temperature of annealing, the nitrogen content of the alloy and the time of the action of hot water or steam, as well as the dependence of the hydrogen content in the alloy on the peripheral tension of the cladding in hot water both in non-active environment and at irradiation with a neutron flux of approximately 1020 n/cm2. (J.B.)

  1. Gradient microstructure in laser clad TiC-reinforced Ni-alloy composite coating

    NARCIS (Netherlands)

    Pei, Y.T.; Zuo, T.C.

    1998-01-01

    A gradient TiC–(Ni alloy) composite coating was produced by one step laser cladding with pre-placed mixture powder on a 1045 steel substrate. The clad layers consisted of TiC particles, γ-Ni primary dendrites and interdendritic eutectics. From the bottom to the top of the clad layer produced at 2000

  2. Prevention of microcracking by REM addition to alloy 690 filler metal in laser clad welds

    International Nuclear Information System (INIS)

    Effect of REM addition to alloy 690 filler metal on microcracking prevention was verified in laser clad welding. Laser clad welding on alloy 132 weld metal or type 316L stainless steel was conducted using the five different filler metals of alloy 690 varying the La content. Ductility-dip crack occurred in laser clad welding when La-free alloy 690 filler metal was applied. Solidification and liquation cracks occurred contrarily in the laser cladding weld metal when the 0.07mass%La containing filler metal was applied. In case of laser clad welding on alloy 132 weld metal and type 316L stainless steel, the ductility-dip cracking susceptibility decreased, and solidification/liquation cracking susceptibilities increased with increasing the La content in the weld metal. The relation among the microcracking susceptibility, the (P+S) and La contents in every weld pass of the laser clad welding was investigated. Ductility-dip cracks occurred in the compositional range (atomic ratio) of La/(P+S) 0.99(on alloy 132 weld metal), >0.90 (on type 316L stainless steel), while any cracks did not occur at La/(P+S) being between 0.21-0.99 (on alloy 132 weld metal) 0.10-0.90 (on type 316L stainless steel). Laser clad welding test on type 316L stainless steel using alloy 690 filler metal containing the optimum La content verified that any microcracks did not occurred in the laser clad welding metal. (author)

  3. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    Science.gov (United States)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  4. Wear resistance and hot corrosion behaviour of laser cladding Co-based alloy

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    2Cr13 stainless steel was surface cladded with Co-based alloy using a high power carbon dioxide laser. The microstructure, wear resistance and corrosion properties of the clad layer were investigated. It is found that the high temperature corrosion behavior and wearing resistant property of the clad layer are 3 and 2.5 times higher than those of the parent metal. Under the high temperature molten lead sulphate salt corrosion condition, the clad layer fails by spalling which is caused by intergrannular corrosion within the clad layer. The fine dendritic structure and the oxide help to retard the penetration of the sulphur ion that induces the intergrannular corrosion.

  5. High burnup effects on the burst behavior of Zr based alloy claddings under LOCA conditions

    International Nuclear Information System (INIS)

    A current loss of coolant accident (LOCA) criterion is based on the results obtained from non pressurized claddings specimens under simulated LOCA condition. However, integrity of fuel cladding can be significantly affected by ballooning and rupture that caused by pressure difference between inner and outer cladding during LOCA. Ballooning may cause the fuel relocation or fuel dispersal due to its rupture opening during accidents. In addition, wall thickness of cladding can be reduced and local regions near the rupture open would become heavily oxidized and hydride d. Therefore, integral test that can simulate whole process during LOCA should be carried out for comprehensive safety analysis. Although a number of researches have been conducted, most investigations of them were performed using as received cladding specimens. In this study, burst behavior of several kinds of zirconium based alloys was investigated by integral LOCA test and high burnup effects on the burst behavior of fuel cladding were also examined using H charged cladding sample

  6. Fuel behavior in severe accidents and Mo-alloy based cladding designs to improve accident tolerance

    International Nuclear Information System (INIS)

    The severe accidents at TMI-2 and Fukushima-Daiichi led to core meltdown and hydrogen explosions. The main source of energy causing core melting is the decay heat from β-, β+, and γ decays of short-lived isotopes following a power scram. The exothermic reaction of Zr-alloy cladding can further increase the cladding temperature leading to rapid cladding corrosion and hydrogen production. The most effective mitigation to minimize core damage in a severe accident is to extend the duration of heat removal capacity via battery-supported passive cooling for as long as practically possible. Replacing the Zr-alloy cladding with a higher heat resistant cladding with lower enthalpy release rate may also provide additional coping time for accident management. Such a heat resistant cladding may also overcome the current licensing concerns about Zr-alloy hydriding and post quench ductility issues in a design base loss of coolant accident (LOCA). Zr-alloy cladding, while has been optimized for normal operation in high pressure water and steam of light water reactors, will rapidly lose its corrosion resistance and tensile and creep strength in high pressure steam. Evaluation of alternate cladding materials and designs have been performed to search for a new fuel cladding design which will substantially improve the safety margins at elevated temperatures during a severe accident, while maintaining the excellent fuel performance attributes of the current Zr-alloy cladding. The screening criteria for the evaluation include neutronic properties, material availability, adaptability and operability in current LWRs, resistance to melting. The new designs also need to be fabricable, maintain sufficient strength and resist to attack by high pressure steam. Engineering metals, alloys and ceramics which can meet some or most of these requirements are limited. Following review of the properties of potential candidates, it is concluded that molybdenum alloys may potentially achieve the

  7. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  8. Laser cladding of Al + Ir powders on ZM5 magnesium base alloy

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Laser cladding of preplaced Al + Ir powders on a ZM5 magnesium alloy was performed to enhance the corrosion resistance of the ZM5 magnesium alloy. A metallurgical bond was obtained at the coating/substrate interface. The corrosion potential (Ecorr) of the laser cladded sample was 169 mV positive to that of the untreated ZM5 substrate, while the corrosion current (Icorr) was some one order of magnitude lower. The laser cladded sample, unlike the untreated ZM5 substrate,showed a passive region in the polarization plot. Immersion tests confirmed that the corrosion resistance of the laser cladded ZM5 sample was significantly enhanced in 3.5 wt.% NaCl solution. The Al-rich phases of AlIr, Mg17Al12, and Al formed in the cladding layer and the rapid solid characteristics were contributed to the improved corrosion behavior of the coating.

  9. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    International Nuclear Information System (INIS)

    Highlights: • Deep surface rolling as a post-repair enhancement technology was applied to the laser cladded 7075-T651 aluminium alloy specimens that simulated corrosion damage blend-out repair. • The residual stresses induced by the deep surface rolling process were measured. • The deep surface rolling process can introduce deep and high magnitude compressive residual stresses beyond the laser clad and substrate interface. • Spectrum fatigue test showed the fatigue life was significantly increased by deep surface rolling. - Abstract: Deep surface rolling can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. To develop cost-effective aircraft structural repair technologies such as laser cladding, deep surface rolling was considered as an advanced post-repair surface enhancement technology. In this study, aluminium alloy 7075-T651 specimens with a blend-out region were first repaired using laser cladding technology. The surface of the laser cladding region was then treated by deep surface rolling. Fatigue testing was subsequently conducted for the laser clad, deep surface rolled and post-heat treated laser clad specimens. It was found that deep surface rolling can significantly improve the fatigue life in comparison with the laser clad baseline repair. In addition, three dimensional residual stresses were measured using neutron diffraction techniques. The results demonstrate that beneficial compressive residual stresses induced by deep surface rolling can reach considerable depths (more than 1.0 mm) below the laser clad surface

  10. Application of the diagrams of phase transformations during aging for optimizing the aging conditions for V1469 and 1441 Al-Li alloys

    Science.gov (United States)

    Lukina, E. A.; Alekseev, A. A.; Antipov, V. V.; Zaitsev, D. V.; Klochkova, Yu. Yu.

    2009-12-01

    To describe the changes in the phase composition of alloys during aging, it is convenient to construct TTT diagrams on the temperature-aging time coordinates in which time-temperature regions of the existence of nonequilibrium phases that form during aging are indicated. As a rule, in constructing the diagrams of phase transformations during aging (DPTA), time-temperature maps of properties are plotted. A comparison of the diagrams with maps of properties allows one to analyze the effect of the structure on the properties. In this study, we analyze the DPTAs of V1469 (Al-1.2 Li-0.46 Ag-3.4 Cu-0.66 Mg) and 1441 (Al-1.8 Li-1.1 Mg-1.6 Cu, C Mg/ C Cu ≈ 1) alloys. Examples of the application of DPTA for the development of steplike aging conditions are reported.

  11. Delayed hydride cracking of zirconium alloy fuel cladding

    International Nuclear Information System (INIS)

    This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

  12. Laser Cladding of Composite Bioceramic Coatings on Titanium Alloy

    Science.gov (United States)

    Xu, Xiang; Han, Jiege; Wang, Chunming; Huang, Anguo

    2016-02-01

    In this study, silicon nitride (Si3N4) and calcium phosphate tribasic (TCP) composite bioceramic coatings were fabricated on a Ti6Al4V (TC4) alloy using Nd:YAG pulsed laser, CO2 CW laser, and Semiconductor CW laser. The surface morphology, cross-sectional microstructure, mechanical properties, and biological behavior were carefully investigated. These investigations were conducted employing scanning electron microscope, energy-dispersive x-ray spectroscopy, and other methodologies. The results showed that both Si3N4 and Si3N4/TCP composite coatings were able to form a compact bonding interface between the coating and the substrate by using appropriate laser parameters. The coating layers were dense, demonstrating a good surface appearance. The bioceramic coatings produced by laser cladding have good mechanical properties. Compared with that of the bulk material, microhardness of composite ceramic coatings on the surface significantly increased. In addition, good biological activity could be obtained by adding TCP into the composite coating.

  13. Microstructure and Wear Behavior of CoCrFeMnNbNi High-Entropy Alloy Coating by TIG Cladding

    OpenAIRE

    2015-01-01

    Alloy cladding coatings are widely prepared on the surface of tools and machines. High-entropy alloys are potential replacements of nickel-, iron-, and cobalt-base alloys in machining due to their excellent strength and toughness. In this work, CoCrFeMnNbNi HEA coating was produced on AISI 304 steel by tungsten inert gas cladding. The microstructure and wear behavior of the cladding coating were studied by X-ray diffraction, scanning electron microscopy, energy dispersive spectrometer, microh...

  14. Laser Cladding of γ-TiAl Intermetallic Alloy on Titanium Alloy Substrates

    Science.gov (United States)

    Maliutina, Iuliia Nikolaevna; Si-Mohand, Hocine; Piolet, Romain; Missemer, Florent; Popelyukh, Albert Igorevich; Belousova, Natalya Sergeevna; Bertrand, Philippe

    2016-01-01

    The enhancement of titanium and titanium alloy's tribological properties is of major interest in many applications such as the aerospace and automotive industry. Therefore, the current research paper investigates the laser cladding of Ti48Al2Cr2Nb powder onto Ti6242 titanium alloy substrates. The work was carried out in two steps. First, the optimal deposition parameters were defined using the so-called "combined parameters," i.e., the specific energy E specific and powder density G. Thus, the results show that those combined parameters have a significant influence on the geometry, microstructure, and microhardness of titanium aluminide-formed tracks. Then, the formation of dense, homogeneous, and defect-free coatings based on optimal parameters has been investigated. Optical and scanning electron microscopy techniques as well as energy-dispersive spectroscopy and X-ray diffraction analyses have shown that a duplex structure consisting of γ-TiAl and α 2-Ti3Al phases was obtained in the coatings during laser cladding. Moreover, it was shown that produced coatings exhibit higher values of microhardness (477 ± 9 Hv0.3) and wear resistance (average friction coefficient is 0.31 and volume of worn material is 5 mm3 after 400 m) compared to those obtained with bare titanium alloy substrates (353 Hv0.3, average friction coefficient is 0.57 and a volume of worn material after 400 m is 35 mm3).

  15. Studies on fuel-clad chemical interaction of U-10Zr alloy with T91 cladding

    International Nuclear Information System (INIS)

    Fuel-clad chemical compatibility has been recognized as one of the major concerns about the performance of metallic fuel since it limits the life of the fuel pin due to formation of low melting eutectic. The fuel-clad compatibility between U-10Zr and T91 was studied by diffusion couple experiments at normal operating and transient conditions. The diffusion reaction between these two was strongly retarded due to formation of a Zr-rich layer at the interface. (author)

  16. An alkali dispenser photocathode (Al-Li)-Ag-O-Li

    International Nuclear Information System (INIS)

    We propose a new Li-actived metallic photocathode that could be used in high-current photo-injectors for linear accelerators. An Ag thin film is evaporated in UHV on the surface of a piece of (Al-Li) alloy. After treating at 380 deg C Li diffuses through Ag and decreases the surface work function to 2.7 eV. The emission yields Qe, in electrons per incident photon, are higher than those measured with (Al-Li) alone. Oxidation of the surface allows one to increase again Qe which reaches Qe∼1.7x10-3 for photon energies of 3.5 and 4.6 eV respectively (energies corresponding to photons delivered by a frequency triples -or quadruples- YAG laser)

  17. Failure analysis of fusion clad alloy system AA3003/AA6xxx sheet under bending

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Y., E-mail: shiyh@mcmaster.ca [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada); Jin, H. [Novelis Global Technology Center, P.O. Box 8400, Kingston, Ontario, Canada K7L 5L9 (Canada); Wu, P.D. [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada); Lloyd, D.J. [Aluminum Materials Consultants, 106 Nicholsons Point Road, Bath, Ontario, Canada K0H 1G0 (Canada); Embury, D. [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada)

    2014-07-29

    An ingot of AA6xxx Al–Si–Mg–Cu alloy clad with AA3003 Al–Mn alloy was co-cast by Fusion technology. Bending tests and numerical modeling were performed to investigate the potential for sub-surface cracking for this laminate system. To simulate particle-induced crack initiation and growth, both random and stringer particles have been selected to mimic the particle distribution in the tested samples. The morphology of cracking in the model was similar to that observed in clad sheet tested in the Cantilever bend test. The crack initiated in the core close to the clad-core interface where the strain in the core is highest, between particles or near particles and propagates along local shear bands in the core, while the clad layer experiences extreme thinning before failure.

  18. Properties and features of structure formation CuCr-contact alloys in electron beam cladding

    Energy Technology Data Exchange (ETDEWEB)

    Durakov, Vasiliy G., E-mail: electron@ispms.tsc.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); Dampilon, Bair V., E-mail: dampilon@ispms.tsc.ru, E-mail: gnusov@rambler.ru; Gnyusov, Sergey F., E-mail: dampilon@ispms.tsc.ru, E-mail: gnusov@rambler.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055, Russia and National Research Tomsk Polytechnic University, Tomsk, 634050 (Russian Federation)

    2014-11-14

    The microstructure and properties of the contact CuCr alloy produced by electron-beam cladding have been investigated. The effect of the electron beam cladding parameters and preheating temperature of the base metal on the structure and the properties of the coatings has been determined. The bimodal structure of the cladding coating has been established. The short circuit currents tests have been carried out according to the Weil-Dobke synthetic circuit simulating procedure developed for vacuum circuit breakers (VCB) test in real electric circuits. Test results have shown that the electron beam cladding (EBC) contact material has better breaking capacity than that of commercially fabricated sintered contact material. The application of the technology of electron beam cladding for production of contact material would significantly improve specific characteristics and reliability of vacuum switching equipment.

  19. Laser Cladding of Magnesium Alloy AZ91D with Silicon Carbide

    Science.gov (United States)

    Cai, L. F.; Mark, C. K.; Zhou, Wei

    Mg alloys are ultralight but their structural applications are often limited by their poor wear and corrosion resistance. The research aimed to address the problem by laser-cladding. Cladding with SiC powder onto surface of AZ91D was carried out using Nd:YAG laser. The laser-clad surface was analyzed using the optical microscope, SEM equipped with EDS, and XRD and found to contain SiC and other Si compounds such as Mg2Si and Al3.21Si0.47 as well as much refined α-Mg grains and β-Mg17Al12 intermetallics. The laser-clad surface possesses considerably higher hardness but its corrosion resistance is not improved, indicating that the laser-cladding technique can only be adopted for applications in noncorrosive environments where wear is the predominant problem.

  20. LASER CLADDING OF MAGNESIUM ALLOY AZ91D WITH SILICON CARBIDE

    OpenAIRE

    L. F. CAI; C. K. MARK; WEI ZHOU

    2009-01-01

    Mg alloys are ultralight but their structural applications are often limited by their poor wear and corrosion resistance. The research aimed to address the problem by laser-cladding. Cladding with SiC powder onto surface of AZ91D was carried out using Nd:YAG laser. The laser-clad surface was analyzed using the optical microscope, SEM equipped with EDS, and XRD and found to contain SiC and other Si compounds such as Mg2Si and Al3.21Si0.47 as well as much refined α-Mg grains and β-Mg17Al12 inte...

  1. Structural Analysis of Surface-Modified Oxidation-Resistant Zirconium Alloy Cladding for Light Water Reactors

    International Nuclear Information System (INIS)

    While the current zirconium-based alloy cladding (Zircaloy, here after) has served well for fission-product barrier and heat transfer medium for the nuclear fuel of light water reactors (LWRs) in steady-states, concerns surrounding its mechanical behavior during accidents have drawn serious attentions. In accidents, strength degradation of the current-zirconium based alloy cladding manifests at temperature around ∼800 .deg. C, which results in fuel ballooning. Above 1000 .deg. C, zircaloy undergoes rapid oxidation with steam. Formation of brittle oxide (ZrO2) and underlying oxygen-saturated α-zircaloy as a consequence of steam oxidation leads to loss of cladding ductility. Indeed, the loss of zircaloy ductility upon the steam oxidation has been taken as a measure of fuel failure criteria as stated in 10 CFR 50.46. In addition, zircaloy steam oxidation is an exothermic reaction, which is an energy source that sharply accelerates temperature increase under loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents (LOCA) before significant fuel/core melting starts. Hydrogen generated as a result of zircaloy oxidation could cause an explosion if certain conditions are met. In steady-state operation, zircaloy embrittlement limits the burnup of the fuel rod to assure remaining cladding ductility to cope with accidents. As a way to suppress hydrogen generation and cladding embrittlement by oxidation, ideas of cladding coating with an oxidation-preventive layer have emerged. Indeed, among a numbers of 'accident-tolerant-fuel (ATF)' concepts, the concept of coating the current fuel rod. Some signs of success on the lab-scale oxidation-prevention have been confirmed with a few coating candidates. Yet, relatively less attention has been given to structural integrity of coated zirconium-based alloy cladding. It is important to note that oxidation-suppression performance

  2. Microstructure of laser clad Ni- Cr- Al- Hf alloy on a γ' strengthened ni- base superalloy

    Science.gov (United States)

    Singh, Jogender; Mazumder, J.

    1988-08-01

    Alloys and coatings for alloys for improved high temperature service life under aggressive atmo-spheres are of great contemporary interest. There is a general consensus that the addition of rare earths such as Hf will provide many beneficial effects for such alloys. The laser cladding technique was used to produce Ni-Cr-AI-Hf alloys with extended solid solution of Hf. A 10 kW CO2 laser with mixed powder feed was used for laser cladding. Optical, scanning electron (SEM) and scanning transmission electron (STEM) microscopy were employed to characterize the microstructure of alloys produced during laser cladding processes. Microstructural studies revealed grain refinement, considerable in-crease in solubility of Hf in the matrix, Hf-rich precipitates, and new metastable phases. The size and morphology of γ' (Ni3Al) phase were discussed in relation to its microchemistry and the laser processing conditions. This paper will report the microstructural development in this laser clad Ni-Cr-AI-Hf alloy.

  3. Laser cladding of ZrO2-(Ni alloy) composite coating

    OpenAIRE

    Pei, Y.T.; Ouyang, J.H.; Lei, T.C.

    1996-01-01

    The microstructure of laser-clad 60 vol.% ZrO2 (partially stabilized with 2 mol% Y2O3) plus 40 vol.% Ni alloy composite coating on steel 1045 was investigated by scanning electron microscopy, electron probe microanalysis, X-ray diffraction, energy-dispersive X-ray analysis and microhardness tests. The composite coating consists of a pure ZrO2 clad layer in the outer region and a bonding zone of Ni alloy adjacent to the substrate. The pure ceramic layer exhibits fine equiaxed ZrO2 grains in th...

  4. Studies of the AA2519 Alloy Hot Rolling Process and Cladding with EN AW-1050A Alloy

    OpenAIRE

    Płonka B.; Rajda M.; Zamkotowicz Z.; Żelechowski J.; Remsak K.; Korczak P.; Szymański W.; Snieżek L.

    2016-01-01

    The objective of the study was to determine the feasibility of plastic forming by hot rolling of the AA2519 aluminium alloy sheets and cladding these sheets with a layer of the EN AW-1050A alloy. Numerous hot-rolling tests were carried out on the slab ingots to define the parameters of the AA2519 alloy rolling process. It has been established that rolling of the AA2519 alloy should be carried out in the temperature range of 400-440°C. Depending on the required final thickness of the sheet met...

  5. Corrosion inhibition of steam generator tubesheet by Alloy 690 cladding in secondary side environments

    Energy Technology Data Exchange (ETDEWEB)

    Hur, Do Haeng, E-mail: dhhur@kaeri.re.kr; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Shim, Hee Sang

    2013-11-15

    Denting is a phenomenon that a steam generator tube is distorted by a volume expansion of corrosion products of the tube support and tubesheet materials adjacent to the tube. Although denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable concern in the crevice region of the top of tubesheet. This paper provides a new technology to prevent denting by cladding the secondary surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material was cladded onto the surface of an SA508 tubesheet to a thickness of about 9 mm. The corrosion rates of the original SA508 tubesheet and the Alloy 690 clad material were measured in acidic and alkaline simulated environments. Using Alloy 690 cladding, the corrosion rate of the tubesheet within a magnetite sludge pile decreased by a factor of 680 in 0.1 M NiCl{sub 2} solution at 300 °C, and by a factor of 58 in 2 M NaOH solution at 315 °C. This means that denting can drastically be prevented by cladding the secondary tubesheet surface with corrosion resistant materials.

  6. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1996-08-01

    In Phase I a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase II (in situ testing) has exposed samples of 347, RA-8511, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, 800HT, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on an air-cooled, retractable corrosion probe, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. Samples of each alloy will be exposed for 4000, 12,000, and 16,000 hours of operation. The results will be presented for the metallurgical examination of the corrosion probe samples after 4000 hours of exposure.

  7. Foam coating on aluminum alloy with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; van Heeswijk, V.; de Hosson, J.T.M.; Csach, K.

    2004-01-01

    dThis article concentrates on the creation of a foam layer on an Al-Si substrate with laser technology. The cladding of At-Si powder in the front of a laser track has been separated from the side injection of mixture of Al-Si/TiH2 powder (foaming agent), which allows for fine tuning of the main proc

  8. Zirconium-niobium alloys as fuel cladding for water cooled reactors

    International Nuclear Information System (INIS)

    Zirconium-niobium alloys containing 1.0 wt% and 2.5 wt% niobium have been investigated for use as fuel cladding. Irradiations were conducted in pressurized and boiling water reactor loops and in a small power reactor. The in-reactor corrosion and hydriding performance of the Zr-2.5 wt% Nb alloy was superior to that of Zircaloy in low oxygen coolant and about the same at higher oxygen levels. Fusion welded end closures performed well but heavy white oxide formed on the beta heated zones; this effect was reduced with a post-weld heat treatment. Delayed hydrogen cracking of resistance-welded end closures was successfully overcome by changing the weld profile and by a post-weld heat treatment. Limited power ramp testing of CANLUB coated fuel elements clad with the Zr-Nb alloys indicates that the tolerance to power ramps is about the same as that of Zircaloy-4 clad fuel with similar coatings. This is somewhat at variance with iodine stress corrosion tests on irradiated cladding which showed that the Zr-Nb alloys were more susceptible to stress corrosion cracking. (author)

  9. Laser cladding of nickel base alloy on SS316L for improved wear and corrosion behaviour

    International Nuclear Information System (INIS)

    Laser cladding by an Nd:YAG laser was employed to deposit Ni base alloy (Ni-Mo-Cr-Si) on stainless steel-316 L substrate. The resulting defect-free clad with minimum dilution of the substrate was characterized by optical microscopy, scanning electron microscopy, X-ray diffraction and Vickers microhardness test. Dry sliding wear of the cladding and the substrate was evaluated using a ball-on-plate reciprocating wear tester against different counter bodies (WC and 52100 Cr steel). The reciprocating sliding wear resistance of the coating was evaluated as a function of the normal load, keeping the sliding amplitude and sliding speed constant. Wear mechanisms were analyzed by observation of wear track morphology using SEM-EDS. The electrochemical corrosion behavior of clad layer was studied in reducing environment (HCl) to estimate the general corrosion resistance of the laser clad layer in comparison with the substrate SS-316L. The clad layer showed higher wear resistance under reducing condition than that of the substrate material stainless steel 316L. (author)

  10. Effect of additives on corrosion resistance of Zirconium alloy for extended burn-up fuel cladding

    International Nuclear Information System (INIS)

    Sumitomo Metal Industries, Ltd. (SMI) supplies Zircaloy cladding tubes and has been developing high corrosion resistance Zr alloys for extended burn-up fuel claddings for BWR and PWR, respectively. For BWR cladding tube, small addition of IVb and Vb elements to Low Sn Zircaloy-2 improved nodular corrosion resistance. It was observed by Transmission Electron Microscopy that these additives complied with a Zr(Cr, Fe)2 type intermetallic compound and those size were finer than that precipitated in a conventional Zircaloy-2. That was assumed to result in suppressing nodular corrosion occurrence. For PWR cladding tube, small addition of Ni and Nb to extremely low Sn zirconium alloy improved uniform corrosion resistance and suppressed hydrogen pick-up. As this results Zr-1.0Sn-0.27Fe-0.16Cr-0.1Nb-0.01Ni were selected as a candidate alloy. In spite of extremely low Sn content, its mechanical properties were almost same as conventional Zircaloy-4. (author)

  11. Out-of-pile performances of new zirconium alloys for PWR fuel cladding

    International Nuclear Information System (INIS)

    Two new zirconium alloys, N18 and N36, containing Sn, Nb, Fe and Cr have been developed to use as superior PWR fuel rod cladding materials. The results are obtained from the out-of-pile performance tests on these advanced alloy claddings or materials. Analytical electron microscopy demonstrated that the best out-of-pile corrosion resistance was obtained for microstructure containing a fine and uniform distribution of β-Nb and/or Zr(Fe, Cr)2 particles. Autoclave testing indicated that N18 and N36 alloys possessed superior corrosion resistance including uniform and nodular corrosion. It has been demonstrated that the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the exposure time or oxide thickness, and hydrogen absorption rate of both alloys is quite low compared to that of Zircaloy-4. These alloys have demonstrated superior out-of-pile tensile strength, burst and creep properties relative to Zircaloy-4. In addition, the thermal physical properties, texture, Stress Corrosion Cracking (SCC) for two new zirconium alloys have been examined, which also showed a good results compared to Zircaloy-4. (author)

  12. Performance and testing of refractory alloy clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO2) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at .% burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  13. White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-05-29

    This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

  14. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  15. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  16. Effect of Specific Energy Input on Microstructure and Mechanical Properties of Nickel-Base Intermetallic Alloy Deposited by Laser Cladding

    Science.gov (United States)

    Awasthi, Reena; Kumar, Santosh; Chandra, Kamlesh; Vishwanadh, B.; Kishore, R.; Viswanadham, C. S.; Srivastava, D.; Dey, G. K.

    2012-12-01

    This article describes the microstructural features and mechanical properties of nickel-base intermetallic alloy laser-clad layers on stainless steel-316 L substrate, with specific attention on the effect of laser-specific energy input (defined as the energy required per unit of the clad mass, kJ/g) on the microstructure and properties of the clad layer, keeping the other laser-cladding parameters same. Defect-free clad layers were observed, in which various solidified zones could be distinguished: planar crystallization near the substrate/clad interface, followed by cellular and dendritic morphology towards the surface of the clad layer. The clad layers were characterized by the presence of a hard molybdenum-rich hexagonal close-packed (hcp) intermetallic Laves phase dispersed in a relatively softer face-centered cubic (fcc) gamma solid solution or a fine lamellar eutectic phase mixture of an intermetallic Laves phase and gamma solid solution. The microstructure and properties of the clad layers showed a strong correlation with the laser-specific energy input. As the specific energy input increased, the dilution of the clad layer increased and the microstructure changed from a hypereutectic structure (with a compact dispersion of characteristic primary hard intermetallic Laves phase in eutectic phase mixture) to near eutectic or hypoeutectic structure (with reduced fraction of primary hard intermetallic Laves phase) with a corresponding decrease in the clad layer hardness.

  17. Laser multi-layer cladding on ZM6 magnesium base alloy

    Institute of Scientific and Technical Information of China (English)

    Changjun Chen(陈长军); Dongsheng Wang(王东生); Maocai Wang(王茂才)

    2003-01-01

    A pulsed Nd: YAG laser is used in multi-layer cladding on ZM6 Mg base alloys. The microstructure isstudied with an optical microscope and a scanning electron microscope (SEM). The composition within thelayer was determined by electron probe microanalysis (EPMA). X-ray diffraction (XRD) was also used toinvestigate the phase of constitutes of the cladding zone. The results show that microstructure in solidifiedcladding layer changes much when treated by high energy laser beam. The microstructure of the ZM6alloy consists of α-Mg and Mg9Nd, while the L-ZM6 of α-Mg, Mg9Nd and c-Zr. The depth of the claddingis over 1 mm. Many fine particles were found to be distributed homogeneously throughout the matrix andthe columnar grain grows along substrate.

  18. Alloy element evaporation phenomenon of low Sn Zr-4 cladding tube during electric-beam welding

    International Nuclear Information System (INIS)

    The author analyses surface constitution for the weld samples of low Sn Zr-4 cladding tube by wave spectrum analytical method. The results show that the chemical constitutions of Sn, Cr and Fe element trends towards statistical increment from peripheral weld to medial weld, and that the constitution of Sn, Cr and Fe elements on surface of the samples being gray corrosion products are sharply lowered. It is obvious that there is a gross evaporation phenomenon for alloy element of the girth welding under certain welding specification while welding of low Sn Zr-4 cladding tube by electric-beam welding, particularly, corrosion resistance of the alloy will be decreased when Sn element evaporation is lower than 0.5% weigh

  19. Interaction between U-9 wt. % Mo fuel and Zr-1 wt. % Nb cladding alloys

    International Nuclear Information System (INIS)

    A few very recent studies have identified the ability of Zr in acting as a diffusion barrier to reduce the deleterious fuel-clad chemical interaction (FCCI), which restricts nuclear fuel designers around the world to successfully utilize the potential of the γ-phase stabilized U-Mo alloys as reduced enrichment fuels in research and test reactor. Further investigations pertaining to metallurgical interaction between U-Mo and Zr are essential not only to establish Zr as a diffusion barrier in U-Mo fuel but also to envisage Zr-base alloys as cladding as against the currently used Al-alloys. In this work, metallurgical interaction between U-9 wt. % Mo metallic fuel alloy and Zr-1 wt.% Nb clad material has been assessed through scanning electron microscopy (SEM), electron probe microanalysis (EPMA) and transmission electron microscopy (TEM). Interdiffusion of constituent elements across the fuel-clad interface, together with the phase reactions occurring at high temperature and during subsequent cooling, resulted in development of a layered interaction zone where coexistence of a bcc solid solution phase with varying compositions, along with α-U, α-Zr and Mo2Zr phases could be noticed. The instability in the γ-U(Mo,Zr) matrix leading to phase separation into α-U and α-Zr phases and the orientation relationships amongst them were established through microdiffraction and composite selected area electron diffraction (SAED) patterns, respectively. The present study is an endeavor to rationalize these observations, which remain unexplained in literature. (author)

  20. Laser Clad ZrO2-Y2O3 Ceramic/Ni-base Alloy Composite Coatings

    OpenAIRE

    Pei, Y.T.; Ouyang, J.H.; Lei, T.C.; Zhou, Y.

    1995-01-01

    A laser cladding technique was used to produce ZrO2-Y2O3 ceramic/Ni-base alloy composite coatings on stainless steel 4Cr13. The microstructure and hardness of the composite coatings are analyzed by XRD, SEM, EPMA, TEM and microhardness testing techniques. A stratification is observed in the laser clad zone. The upper region of the clad is a pure ZrO2 ceramic layer, and the lower region is an excellent transition layer of Ni-base alloy. The ZrO2 ceramic layer exhibits equiaxed grains and colum...

  1. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Krawchuk, M.T.; Van Weele, S.F. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1995-08-01

    A number of developmental and commercial tubing alloys and claddings have previously been exposed in Phase I to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. This program is exposing samples of TP 347, RA-85H, HR-3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF-709, 690 clad, and 671 clad, which showed good corrosion resistance from Phase 1, to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and are being controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The exposure will continue for 4000, 12,000, and 16,000 hours of operation. After the three exposure times, the samples will be metallurgically examined to determine the wastage rates and mode of attack. The probes were commissioned November 16, 1994. The temperatures are being recorded every 15 minutes, and the weighted average temperature calculated for each sample. Each of the alloys is being exposed to a temperature in each of two temperature bands-1150 to 1260{degrees}F and 1260 to 1325{degrees}F. After 2000 hours of exposure, one of the corrosion probes was cleaned and the wall thicknesses were ultrasonically measured. The alloy performance from the field probes will be discussed.

  2. Laser cladding of a Mg based Mg-Gd-Y-Zr alloy with Al-Si powders

    Science.gov (United States)

    Chen, Erlei; Zhang, Kemin; Zou, Jianxin

    2016-03-01

    In the present work, a Mg based Mg-Gd-Y-Zr alloy was subjected to laser cladding with Al-Si powders at different laser scanning speeds in order to improve its surface properties. It is observed that the laser clad layer mainly contains Mg2Si, Mg17Al12 and Al2(Gd,Y) phases distributed in the Mg matrix. The depth of the laser clad layer increases with decreasing the scanning speed. The clad layer has graded microstructures and compositions. Both the volume fraction and size of Mg2Si, Mg17Al12 and Al2(Gd,Y) phases decreases with the increasing depth. Due to the formation of these hardening phases, the hardness of clad layer reached a maximum value of HV440 when the laser scanning speed is 2 mm/s, more than 5 times of the substrate (HV75). Besides, the corrosion properties of the untreated and laser treated samples were all measured in a NaCl (3.5 wt.%) aqueous solution. The corrosion potential was increased from -1.77 V for the untreated alloy to -1.13 V for the laser clad alloy with scanning rate of 2 mm/s, while the corrosion current density was reduced from 2.10 × 10-5 A cm-2 to 1.64 × 10-6 A cm-2. The results show that laser cladding is an efficient method to improve surface properties of Mg-Rare earth alloys.

  3. Influence of Zirconia on Hydroxyapatite Coating on Ti-Alloy by Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    杜海燕; 霍伟荣; 高海; 王丽娟; 邱世鹏; 刘家臣

    2003-01-01

    Coating titanium alloy with the bioceramic material hydroxyapatite(HAP) has been used to improve the poor osteoinductive properties of pure titanium alloy. But in clinical applications, the mechanical failure of HAP-coated titanium alloy implant suffered at the interface of the HAP coatings and titanium alloy substrate will be a potential weakness in prosthesis. Yttria-stablized zirconia (YSZ) is expected to enhance the mechanical properties of the HAP coating and reduce the coefficient of thermal expansion difference between the coated layer and the substrate. These may reinforce the bonding strength between the coatings and the substrate. In this paper, HAP/YSZ composite coatings were cladded by laser. The effects of zirconia on the microstructure, mechanical properties and formation of tricalcium phosphate (TCP, Ca3(PO4)2) of the HAP/YSZ composite coatings were evaluated. XRD, SEM and TEM were used to investigate the phase composition, microstructure and morphology of the coatings. The experimental results showed that adding YSZ in coatings was favorable to the composition and stability of HAP, and to the improvement of the adhesion strength, microhardness and microtoughness. A well uniform, crack-free coating of HAP/YSZ composites was formed on Ti-alloy substrate by laser cladding.

  4. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  5. Advanced ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N [ORNL; Unocic, Kinga A [ORNL; Hoelzer, David T [ORNL; Pint, Bruce A [ORNL

    2014-09-01

    ODS FeCrAl alloys are being developed with optimum composition and properties for accident tolerant fuel cladding. Two oxide dispersion strengthened (ODS) Fe-15Cr-5Al+Y2O3 alloys were fabricated by ball milling and extrusion of gas atomized metallic powder mixed with Y2O3 powder. To assess the impact of Mo on the alloy mechanical properties, one alloy contained 1%Mo. The hardness and tensile properties of the two alloys were close and higher than the values reported for fine grain PM2000 alloy. This is likely due to the combination of a very fine grain structure and the presence of nano oxide precipitates. The nano oxide dispersion was however not sufficient to prevent grain boundary sliding at 800 C and the creep properties of the alloys were similar or only slightly superior to fine grain PM2000 alloy. Both alloys formed a protective alumina scale at 1200 C in air and steam and the mass gain curves were similar to curves generated with 12Cr-5Al+Y2O3 (+Hf or Zr) ODS alloys fabricated for a different project. To estimate the maximum temperature limit of use for the two alloys in steam, ramp tests at a rate of 5 C/min were carried out in steam. Like other ODS alloys, the two alloys showed a significant increase of the mas gains at T~ 1380 C compared with ~1480 C for wrought alloys of similar composition. The beneficial effect of Yttrium for wrought FeCrAl does not seem effective for most ODS FeCrAl alloys. Characterization of the hardness of annealed specimens revealed that the microstructure of the two alloys was not stable above 1000 C. Concurrent radiation results suggested that Cr levels <15wt% are desirable and the creep and oxidation results from the 12Cr ODS alloys indicate that a lower Cr, high strength ODS alloy with a higher maximum use temperature could be achieved.

  6. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  7. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1998-06-01

    In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

  8. Laser cladding of tungsten carbides (Spherotene ®) hardfacing alloys for the mining and mineral industry

    Science.gov (United States)

    Amado, J. M.; Tobar, M. J.; Alvarez, J. C.; Lamas, J.; Yáñez, A.

    2009-03-01

    The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene ® powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase ®). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed.

  9. Laser cladding of tungsten carbides (Spherotene) hardfacing alloys for the mining and mineral industry

    International Nuclear Information System (INIS)

    The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed

  10. Amorphous structure in a laser clad Ni-Cr-Al coating on Al-Si alloy

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    A mixing microstructure containing Ni-based amorphous structures was observed by TEM in the laser cladzones. As the uniformity of chemical composition and temperature is poor in the laser cladding, the amorphous structurewith some Ni3Al crystals coexists in the cladding. The microhardness of the mixing amorphous structure is HV 600 ~800, which is lower than that of crystal phases in the coating. Differential thermal analysis (DTA) shows that Ni-basedamorphous structure exhibits a higher initial crystallizing temperature (about 588 ℃ ), which is slightly higher than that ofthe eutectic temperature of Al-Si alloy. The wear test results indicate that there are some amorphous structures in the laserclad coating, which reduces the peeling of the granular phases from matrix, and improves the wear resistance

  11. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author)

  12. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    International Nuclear Information System (INIS)

    U3Si2-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  13. Amorphous Structures in Laser Cladding of ZL111 Aluminum Alloy:Semi-quantitative Study by Differential Thermal Analysis (DTA)

    Institute of Scientific and Technical Information of China (English)

    LI Xianqin; CHENG Zhaogu; XIA Jin'an; XU Guoliang; LIANG Gongying

    2000-01-01

    This paper deals with amorphous structures in the laser cladding. ZL111 alloy is the substrate and Ni-Cr-Al alloy is sprayed on the substrate as the coating material. The coating is clad by a 5 kW transverse flow CO2 laser. The observation of SEM and TEM reveal that in the laser cladding there are amorphous structures of two different morphologies: one is space curved flake-like, and exists in the white web-like structures; the other is fir leaf-like, and exists in the grain-like structures. Differential thermal analysis (DTA) is used to semi-quantitatively determine the content of the amorphous structures. A relation is obtained between the content of amorphous structures and the dimensionless laser cladding parameter C. We also show the changes of the amorphous structures after annealing.

  14. Tribological properties of laser cladding TiB2 particles reinforced Ni-base alloy composite coatings on aluminum alloy

    Institute of Scientific and Technical Information of China (English)

    Long He; Ye-Fa Tan; Xiao-Long Wang; Qi-Feng Jing; Xiang Hong

    2015-01-01

    To improve the wear resistance of aluminum alloy frictional parts,TiB2 particles reinforced Ni-base alloy composite coatings were prepared on aluminum alloy 7005 by laser cladding.The microstructure and tribological properties of the composite coatings were investigated.The results show that the composite coating contains the phases of NiAl,Ni3Al,Al3Ni2,TiB2,TiB,TiC,CrB,and Cr23C6.Its microhardness is HV0.5 855.8,which is 15.4 % higher than that of the Ni-base alloy coating and is 6.7 times as high as that of the aluminum alloy.The friction coefficients of the composite coatings are reduced by 6.8 %-21.6 % and 13.2 %-32.4 % compared with those of the Ni-base alloy coatings and the aluminum alloys,while the wear losses are 27.4 %-43.2 % less than those of the Ni-base alloy coatings and are only 16.5 %-32.7 % of those of the aluminum alloys at different loads.At the light loads ranging from 3 to 6 N,the calculated maximum contact stress is smaller than the elastic limit contact stress.The wear mechanism of the composite coatings is micro-cutting wear,but changes into multi-plastic deformation wear at 9 N due to the higher calculated maximum contact stress than the elastic limit contact stress.As the loads increase to 12 N,the calculated flash temperature rises to 332.1 ℃.The composite coating experiences multi-plastic deformation wear,micro-brittle fracture wear,and oxidative wear.

  15. Microstructure and Wear Behavior of CoCrFeMnNbNi High-Entropy Alloy Coating by TIG Cladding

    Directory of Open Access Journals (Sweden)

    Wen-yi Huo

    2015-01-01

    Full Text Available Alloy cladding coatings are widely prepared on the surface of tools and machines. High-entropy alloys are potential replacements of nickel-, iron-, and cobalt-base alloys in machining due to their excellent strength and toughness. In this work, CoCrFeMnNbNi HEA coating was produced on AISI 304 steel by tungsten inert gas cladding. The microstructure and wear behavior of the cladding coating were studied by X-ray diffraction, scanning electron microscopy, energy dispersive spectrometer, microhardness tester, pin-on-ring wear tester, and 3D confocal laser scanning microscope. The microstructure showed up as a nanoscale lamellar structure matrix which is a face-centered-cubic solid solution and niobium-rich Laves phase. The microhardness of the cladding coating is greater than the structure. The cladding coating has excellent wear resistance under the condition of dry sliding wear, and the microploughing in the worn cladding coating is shallower and finer than the worn structure, which is related to composition changes caused by forming the nanoscale lamellar structure of Laves phase.

  16. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1997-12-01

    In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

  17. High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

    International Nuclear Information System (INIS)

    Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

  18. High-temperature steam oxidation kinetics of the E110G cladding alloy

    Science.gov (United States)

    Király, Márton; Kulacsy, Katalin; Hózer, Zoltán; Perez-Feró, Erzsébet; Novotny, Tamás

    2016-07-01

    In the course of recent years, several experiments were performed at MTA EK (Centre for Energy Research, Hungarian Academy of Sciences) on the isothermal high-temperature oxidation of the improved Russian cladding alloy E110G in steam/argon atmosphere. Using these data and designing additional supporting experiments, the oxidation kinetics of the E110G alloy was investigated in a wide temperature range, between 600 °C and 1200 °C. For short durations (below 500 s) or high temperatures (above 1065 °C) the oxidation kinetics was found to follow a square-root-of-time dependence, while for longer durations and in the intermediate temperature range (800-1000 °C) it was found to approach a cube-root-of-time dependence rather than a square-root one. Based on the results a new best-estimate and a conservative oxidation kinetics model were created.

  19. Chemical compatibility between U-6wt.%Zr alloy and T91 cladding

    International Nuclear Information System (INIS)

    Full text: Metal fuel based on binary U-Pu and ternary U-Pu-Zr alloys has been considered as a promising advanced fuel for fast reactor in India due to its favorable thermal and neutronic performance, enhanced reactor safety, ease of fabrication and suitability for pyro-reprocessing. Fuel-clad chemical interaction (FCCI) has been recognized as one of the major issues in metallic fuel since it limits life of fuel pin due to formation of low melting eutectic at significantly lower temperature than the melting point of the fuel alloy. As a part of metallic fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded ternary U-Pu-Zr (Zr2 type layer of thickness around 5μm on the clad side, a Zr depleted layer of thickness 20 μm on the fuel side and a thin Zr rich layer of thickness 2-3μm between the above two. The Zr rich layer formed at the interface is known to be fuel-clad diffusion barrier. U at the interface reacts with Fe, Cr of T91 leaving behind Zr which forms a barrier layer at the interface. When the annealing temperature was increased to 750 deg C, a phase with eutectic microstructure was observed. The bright phase is U6Fe phase, the grey and dark phases are U(Fe,Cr)2 and Zr(Fe,Cr)2 respectively. The results indicate that measured eutectic temperature between U-6Zr and T91 is very crucial from reactor safety point of view. Diffusion couple experiments clearly show that Zr rich layer will retard the diffusion of U to the interface below eutectic temperature but above eutectic temperature U diffusion is fast enough to result in eutectic me

  20. Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors

    Institute of Scientific and Technical Information of China (English)

    B.X. Zhou; M. Y. Yao; Z.K. Li; X.M. Wang; J. Zhoua; C.S. Long; Q. Liu; B.F. Luan

    2012-01-01

    In order to optimize the microstructure and composition of N18 zirconium alloy (Zr-1Sn-0.35Nb-0.35Fe-0.1Cr, in mass fraction, %), which was developed in China in 1990s, the effect of microstructure and composition variation on the corrosion resistance of the N18 alloy has been investigated. The autoclave corrosion tests were carried out in super heated steam at 400 ~C/10.3 MPa, in deionized water or lithiated water with 0.01 mol/L LiOH at 360 ~C/18.6 MPa. The exposure time lasted for 300-550 days according to the test temperature. The results show that the microstructure with a fine and uniform distribution of second phase particles (SPPs), and the decrease of Sn content from 1% (in mass fraction, the same as follows) to 0.8% are of benefit to improving the corrosion resistance; It is detrimental to the corrosion resistance if no Cr addition. The addition of Nb content with upper limit (0.35%) is beneficial to improving the corrosion resistance. The addition of Cu less than 0.1% shows no remarkable influence upon the corrosion resistance for N18 alloy. Comparing the corrosion resistance of the optimized N18 with other commercial zirconium alloys, such as Zircaloy-4, ZIRLO, E635 and Ell0, the former shows superior corrosion resistance in all autoclave testing conditions mentioned above. Although the data of the corrosion resistance as fuel cladding for high burn-up has not been obtained yet, it is believed that the optimized N18 alloy is promising for the candidate of fuel cladding materials as high burn-up fuel assemblies. Based on the theory that the microstructural evolution of oxide layer during corrosion process will affect the corrosion resistance of zirconium alloys, the improvement of corrosion resistance of the N18 alloy by obtaining the microstructure with nano-size and uniform distribution of SPPs, and by decreasing the content of Sn and maintaining the content of Cr is discussed.

  1. Laser Cladding of an Al-11.7Wt% Si Alloy on ZM5 Magnesium Alloy to Enhance the Corrosion Resistance

    Institute of Scientific and Technical Information of China (English)

    CHEN Chang-jun; WANG Mao-cai; WANG Dong-sheng

    2004-01-01

    Magnesium alloy is an important engineering materials, but the wider application is restricted by poor corrosion resistance. An attempt was made to enhance the corrosion resistance and microhardness of a Mg-Al base ZM5 alloy by laser cladding of Al-11.7Wt%Si alloy powder with thickness 1.1mm and 1.7mm. The microstructure, phase and corrosion properties were analyzed by scanning electron micrographic (SEM), electron probe microanalysis(EPMA), vicker hardness tester and corrosion measurement system, respectively. Microhardness of the cladding layer was enhanced to 150-375Hv as compared to 60-99Hv of the substrate. The corrosion potential (Ecorr) of the cladding sample was 80mv higher than the substrate, while the corrosion current (Icorr) was lower than the substrate.

  2. Laser Cladding of an Al-11.7Wt% Si Alloy on ZM5 Magnesium Alloy to Enhance the Corrosion Resistance

    Institute of Scientific and Technical Information of China (English)

    CHENChang-jun; WANGMao-cai; WANGDong-sheng

    2004-01-01

    Magnesium alloy is an important engineering materials, but the wider application is restricted by poor corrosion resistance. An attempt was made to enhance the corrosion resistance and microhardness of a Mg-Al base ZM5 alloy by laser cladding of A1-11.7Wt%Si alloy powder with thickness 1.1 mm and 1.7inm. The microstructure, phase and corrosion properties were analyzed by scanning electron micrographic (SEM), electron probe microanalysis(EPMA), vicker hardness tester and corrosion measurement system, respectively. Microhardness of the cladding layer was enhanced to 150-375Hv as compared to 60-99Hv of the substrate. The corrosion potential (Ecorr) of the cladding sample was 80mv higher than the substrate, while the corrosion current (lcorr) was lower than the substrate.

  3. Development of cladding materials composed of alloys with high compatibility to each corrosive environment on pressure boundaries in nuclear plants

    International Nuclear Information System (INIS)

    Pressure boundary materials used in severe corrosive nuclear environments were developed by means of new alloy designs for attaining the sufficient thermodynamical stability against both heavy irradiations and chemical attacks. Type F5 stainless steel with high austenite phase stability and nickel base silicide dispersed alloy so-called the HWI alloy with the high wear corrosion resistance were developed for core materials in water cooling type nuclear reactors. Three kind of alloys, namely, type 304ULC(EB-SAR), nickel base Cr-W-Si alloy so-called the RW alloy and niobium base alloys which have each different oxidation potential region on these application were developed for vessel materials used in nitric acid environments on reprocessing plants of spent nuclear fuels. The corrosion resistance and the workability of these alloys were improved markedly by means of the electron beam melting for removing harmful impurities in alloy matrixes and the thermomechanical treatment so-called SAR for modifying micro-structures. For improving all-round properties required for pressure boundary materials, cladding technologies between corrosion resistant materials and heat resistant materials were developed by means of diffusion bonding and hydro-isostatic pressing. These cladding process were optimized by both experimentally and theoretically. (author)

  4. Role of Laser Cladding Parameters in Composite Coating (Al-SiC) on Aluminum Alloy

    Science.gov (United States)

    Riquelme, Ainhoa; Escalera-Rodriguez, María Dolores; Rodrigo, Pilar; Rams, Joaquin

    2016-08-01

    The effect of the different control parameters on the laser cladding fabrication of Al/SiCp composite coatings on AA6082 aluminum alloy was analyzed. A high-power diode laser was used, and the laser control parameters were optimized to maximize the size (height and width) of the coating and the substrate-coating interface quality, as well as to minimize the melted zone depth. The Taguchi DOE method was applied using a L18 to reduce the number of experiments from 81 to only 18 experiments. Main effects, signal-noise ratio and analysis of variance were used to evaluate the effect of these parameters in the characteristics of the coating and to determine their optimum values. The influence of four control parameters was evaluated: (1) laser power, (2) scanning speed, (3) focal condition, and (4) powder feed ratio. Confirmation test with the optimal control parameters was carried out to evaluate the Taguchi method's effectivity.

  5. Amorphous structure and properties in laser-clad Ni-Cr-Al coating on Al-Si alloy

    Science.gov (United States)

    Liang, Gongying; Wong, T. T.; Su, J. Y.; Woo, C. H.

    1999-09-01

    A Ni-Cr-Al coating was clad by a 5 kW CO2 laser with different laser power on Al-Si alloy. Using transmission electron microscopy, a mixing microstructure containing Ni- based amorphous structures was observed in the laser clad zones. As the uniformity of chemical composition and temperature is poor in the laser cladding, the amorphous structure with some Ni3Al crystals coexisted in the cladding. According to the morphologies of Ni-based amorphous structures, the amorphous structure existed not only in the net-like boundaries surrounding the granular structure but also in the granular structure. The microhardness of the mixture amorphous structure is between HV 600 - 800, which is lower than that of crystal phases in the coating. A differential thermal analysis showed that Ni- based amorphous structure exhibits a higher initial crystallizing temperature (about 588 degree(s)C), which is slightly higher than that of the eutectic temperature of Al- Si alloy. The wear experimental results showed that some amorphous structure exist in the laser cladding can reduce the peeling of the granular phases from matrix, and improve the its wear resistance.

  6. Fiber laser cladding of nickel-based alloy on cast iron

    Science.gov (United States)

    Arias-González, F.; del Val, J.; Comesaña, R.; Penide, J.; Lusquiños, F.; Quintero, F.; Riveiro, A.; Boutinguiza, M.; Pou, J.

    2016-06-01

    Gray cast iron is a ferrous alloy characterized by a carbon-rich phase in form of lamellar graphite in an iron matrix while ductile cast iron presents a carbon-rich phase in form of spheroidal graphite. Graphite presents a higher laser beam absorption than iron matrix and its morphology has also a strong influence on thermal conductivity of the material. The laser cladding process of cast iron is complicated by its heterogeneous microstructure which generates non-homogeneous thermal fields. In this research work, a comparison between different types of cast iron substrates (with different graphite morphology) has been carried out to analyze its impact on the process results. A fiber laser was used to generate a NiCrBSi coating over flat substrates of gray cast iron (EN-GJL-250) and nodular cast iron (EN-GJS-400-15). The relationship between processing parameters (laser irradiance and scanning speed) and geometry of a single laser track was examined. Moreover, microstructure and composition were studied by Scanning Electron Microscopy (SEM), Energy Dispersive X-Ray Spectroscopy (EDS) and X-Ray Diffraction (XRD). The hardness and elastic modulus were analyzed by means of micro- and nanoindentation. A hardfacing coating was generated by fiber laser cladding. Suitable processing parameters to generate the Ni-based alloy coating were determined. For the same processing parameters, gray cast iron samples present higher dilution than cast iron samples. The elastic modulus is similar for the coating and the substrate, while the Ni-based coating obtained presents a significantly superior hardness than cast iron.

  7. The Role of X-Ray Diffraction for Analyzing Zr-Sn-Nb-Fe Alloys as Power Reactor Fuel Cladding

    OpenAIRE

    Sugondo

    2010-01-01

    Synthesis of Zr-1%Nb-1%Sn-1%Fe alloy is undertaken in order to develop fuel cladding alloy at high burn-up. Powder specimens of Zr-Sn-Nb-Fe alloy were prepared and then formed into pellets with a dimension of 10 mm in height 10 mm in diameter using a pressure of 1.2 ton/cm2. The 5 gram green pellets were then melted in an arc furnace crucible under argon atmosphere. The pressure in the furnace was set at 2 psi and the current was 50 A. Afterwards, the ingots were heated at a temperature of 11...

  8. Study of the degradation mechanisms of zirconium alloy nuclear fuel claddings in air at high-temperature

    International Nuclear Information System (INIS)

    In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reactor (PWR) fuel assemblies: air ingress following a breach in the reactor vessel, de-flooding during handling, spent fuel storage pool de-flooding. Deprived of cooling source, the assemblies temperature raises and the fuel cladding, made out of zirconium based alloys, oxidize. Compared to a steam oxidation, the degradation kinetic of the cladding is higher, on the one hand because of the high enthalpy of the zirconium-oxygen reaction (compared to zirconium-steam reaction), on the other hand because of the nitrogen contribution to the degradation. Temperature escalation and reaction runaway are expected and can rapidly lead to the loss of integrity of the cladding tubes. The objective of this PhD thesis was to affine the understanding of the high temperature air oxidation mechanisms of the two mostly used zirconium alloys in French PWR, Zircaloy-4 and M5. Special attention has been paid to clarify the role of nitrogen. As-received Zircaloy-4 and M5 claddings segments have been oxidized in a thermo-balance in air in isothermal conditions at temperatures between 800 C and 1000 C. Several characterization techniques (micro-Raman spectroscopy, EPMA, XRD, optical and scanning electron microscopies...) have been used to analyze the oxide layers. Identification and evolution of the different phases (monoclinic, tetragonal and cubic zirconia, zirconium oxynitride and ZrN) has been evidenced and analyzed at several step of the oxidation process. Oxidation mechanisms have been proposed and the better oxidation resistance of the M5 alloy, compared to Zircaloy-4 alloy, has been explained. The collected information will allow improvement of modeling aiming to predict the behavior of the claddings in various accidental situations with air ingress (temperature transients, evolution of the gas phase composition...). (author)

  9. A Prediction Study on Oxidation of Aluminum Alloy Cladding of U3Si2-Al Fuel Plate

    International Nuclear Information System (INIS)

    U3Si2-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

  10. High-temperature interaction of fuel rod cladding material (Zr1%Nb alloy) with oxygen-containing mediums

    International Nuclear Information System (INIS)

    The experimental data on kinetics of Zr1%Nb alloy oxidation in steam at atmospheric pressure in the temperature range 550 to 1600 deg. C are presented. The effect of fuel rod claddings deformation on zirconium alloy interaction with steam is shown. The estimates of influence of the additives of air, nitrogen and hydrogen in mixtures with steam at atmospheric pressure on kinetics of steam/zirconium reaction in the temperature range 800 to 1200 deg. C are presented. The correlations for determination of weight gain with indication of area of applicability in space of parameters (temperature, time, deformation, pressure) are shown. (author). 10 refs, 10 figs

  11. Effect of yttrium additions on void swelling in Liquid Metal Fast Breeder Reactor candidate cladding alloys

    International Nuclear Information System (INIS)

    Candidate Liquid Metal Fast Breeder Reactor cladding alloys AL1 (Fe-26% Ni-9% Cr) and AL2 (Fe-35% Ni-12% Cr) without and with the addition of 0.1% yttrium were bombarded by 4 MeV56Fe2+ ions without and with simultaneous bombardment by 0.4 MeV 4He+ ions. These bombardments were conducted at various irradiation temperatures to determine the effect of yttrium on void swelling. The addition of yttrium decreased peak swelling for 4 MeV 56Fe2+ ion bombarded AL1 and AL2 by 28% and 20%, respectively. In all cases where similar sample comparisons were made (i.e., undoped with undoped and doped with doped) and where bombardment conditions were similar (i.e., single with single beam and dual with dual beam), AL1 showed less peak swelling than did AL2. Simultaneously implanting helium during heavy-ion bombardment increased peak swelling in undoped and doped AL1 by factors of 2.3 and 2.6, respectively

  12. Corrosion of aluminum-clad alloys in wet spent fuel storage

    International Nuclear Information System (INIS)

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

  13. Hydrogen absorption by zirconium alloy cladding tube with surface oxide film

    International Nuclear Information System (INIS)

    Hydrogen absorption kinetics of Zircaloy (2 and 4) and Zr-Nb (1% and 2.5%) cladding tubes were studied by heating in hydrogen gas after oxide film formation in steam, oxygen or air. Hydrogen absorption rate depended on the degree of pre-oxidation. In Zr-Nb, the absorption rate was also sensitive to the atmosphere used for pre-oxidation, whereas in Zircaloy the rate was relatively independent of the kind of oxidant. In all materials, pre-oxidation to the transition point was found to bring about high absorption rate in the subsequent hydriding step. After pre-oxidation to the post-transition region, hydrogen absorption rate by Zircaloy showed constant or slightly decreasing tendency with increasing oxidation level, whereas in Zr-Nb, particularly in Zr-2.5%Nb, the rate showed a clearly decreasing tendency depending on the pre-oxidation atmosphere. Different characteristics of Zircaloy and Zr-Nb can partly be explained in terms of different valencies of alloying elements which influence the lattice defect concentrations in the oxide films. (author)

  14. Process modifications in the manufacture of zirconium alloy clad tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Full text: Indian Nuclear Power Reactors use fuel bundles containing Uranium di- oxide pellets clad in Zirconium alloy tubes. The fuel tubes required for these fuel bundles are subjected to severe irradiation and corrosive environment in the nuclear reactor in addition to the exposure of high temperatures. Hence they are manufactured conforming to strict specifications with respect to mechanical, chemical, metallurgical properties and close dimensional tolerances. The fuel tubes are normally manufactured by processing the extruded blanks using 3 or 4 stage cold pilgering with intermediate heat treatment and final finishing operations. This presentation gives a brief detail of process modifications made in the manufacture of these tubes, to improve the mechanical and metallurgical properties and also to improve the overall material recovery. It was established that the variation in wall thickness in the extruded blanks is carried forward to subsequent passes which affects the mechanical properties of the tubes. The annealing parameters were optimized to achieve finer grain size, resulting in better mechanical properties Some of the important modifications carried out include the following: 1. Reducing the wall thickness variation and improving the surface finish of extruded blanks by machining, 2. Introduction of ultrasonic testing of blanks, 3. Optimization of parameters at all stages of pilgering i.e., blank, intermediate and final pass stages, 4. Optimization of parameters at annealing and sand blasting operations, and 5. Improvisation of finishing operations

  15. Prediction model for oxide thickness on aluminum alloy cladding during irradiation

    International Nuclear Information System (INIS)

    An empirical model predicting the oxide film thickness on aluminum alloy cladding during irradiation has been developed as a function of irradiation time, temperature, heat flux, pH, and coolant flow rate. The existing models in the literature are neither consistent among themselves nor fit the measured data very well. They also lack versatility for various reactor situations such as a pH other than 5, high coolant flow rates, and fuel life longer than ∼1200 hrs. Particularly, they were not intended for use in irradiation situations. The newly developed model is applicable to these in-reactor situations as well as ex-reactor tests, and has a more accurate prediction capability. The new model demonstrated with consistent predictions to the measured data of UMUS and SIMONE fuel tests performed in the HFR, Petten, tests results from the ORR, and IRIS tests from the OSIRIS and to the data from the out-of-pile tests available in the literature as well. (author)

  16. Microstructure And Oxidation Properties Of Laser Clad Ni70AL20Cr7Hf3 Alloys With Extended Solid Solution Of Hf

    Science.gov (United States)

    Mazumder, J.; Sircar, S.; Ribaudo, C.; Kar, A.,

    1989-01-01

    Alloys coatings for superalloys for improved higher temperature (1200°C) service life under aggressive atmospheres are of great interest at present. There is a general consensus that addition of rare earths such as hafnium (Hf) to these alloys has a pronounced effect on the oxidation resistance properties at high temperatures. In situ laser cladding technique was used to produce Ni-Al-Cr-Hf alloys with extended solid solution of Hf in a near stoichiometric Ni3Al matrix. A 10 kW CW CO2 laser was used in conjunction with a screw-feed powder dispenser to perform the in situ cladding process.

  17. Temperature dependences of the delayed hydride cracking rate of fuel claddings made of zirconium alloys of various compositions

    Science.gov (United States)

    Markelov, V. A.; Gusev, A. Yu.; Kotov, P. V.; Novikov, V. V.; Saburov, N. S.

    2014-04-01

    The temperature dependences of the delayed hydride cracking (DHC) rate of Zr-1Nb and Zr-0.8Nb-0.8Sn-0.3Fe alloy claddings are studied in the range 127-300°C in comparison with the data obtained for Zr-2.5Nb and Zircaloy-4 alloys earlier. The samples are in the state of cold deformation and stress relief at 400°C for 24 h and in the state of preliminary hydrogen saturation to a hydrogen concentration of 0.02 wt %. As the strength of a zirconium alloy decreases and its ductility increases, the DHC rate and its high-temperature limit for a linear Arrhenius equation decreases, and the fractographic patterns of the fracture surfaces are different.

  18. Laser cladding of Zr-based coating on AZ91D magnesium alloy for improvement of wear and corrosion resistance

    Indian Academy of Sciences (India)

    Kaijin Huang; Xin Lin; Changsheng Xie; T M Yue

    2013-02-01

    To improve the wear and corrosion resistance of AZ91D magnesium alloy, Zr-based coating made of Zr powder was fabricated on AZ91D magnesium alloy by laser cladding. The microstructure of the coating was characterized by XRD, SEM and TEM techniques. The wear resistance of the coating was evaluated under dry sliding wear test condition at room temperature. The corrosion resistance of the coating was tested in simulated body fluid. The results show that the coating mainly consists of Zr, zirconium oxides and Zr aluminides. The coating exhibits excellent wear resistance due to the high microhardness of the coating. The main wear mechanism of the coating and the AZ91D sample are different, the former is abrasive wear and the latter is adhesive wear. The coating compared to AZ91D magnesium alloy exhibits good corrosion resistance because of the good corrosion resistance of Zr, zirconium oxides and Zr aluminides in the coating.

  19. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  20. A new model of hydrogen redistribution in Zr alloy claddings during waterside corrosion in a temperature gradient

    Science.gov (United States)

    Veshchunov, M. S.; Shestak, V. E.; Ozrin, V. D.

    2016-04-01

    A new model for hydrogen spatial redistribution and hydride precipitation in Zr alloys during waterside corrosion extends the traditional approach, valid for consideration of a relatively low volume fraction of the precipitated hydride phase, to a more general case of heavily precipitated hydrides typical for high-burnup fuel cladding tubes of pressurized water reactors and also observed in various autoclave corrosion tests with high hydrogen supercharging. Being implemented in the SVECHA/QUENCH (S/Q) code, the new model reasonably explains various observations in corrosion tests at constant temperature and under temperature gradient as well as under in-reactor corrosion conditions.

  1. Study on laser-cladding Ni-Al-WC alloy layer on the surface of chrome cast iron and alloy layer's micro-structure and properties

    International Nuclear Information System (INIS)

    Laser-cladding Ni-Al-WC alloy layer on the surface of chrome cast iron and alloy layer's micro-structure and properties are studied. The chemical composition, the phase structure, the average micro-hardness, the wear resistance and the corrosion resistance are analyzed for the Ni-Al-WC and the matrix, respectively. The results show that the metallurgical combination is achieved between the spray alloy layer and the surface of chrome cast iron, the chemical composition and micro-structure in the surface layer of the specimen are changed basically, and the micro-hardness, the wear resistance, the corrosion resistance in the surface layer are increased with a large range

  2. Mechanical behavior of Al-Li-SiC composites: Part I. Microstructure and tensile deformation

    Science.gov (United States)

    Poza, P.; Llorca, J.

    1999-03-01

    The microstructure and tensile properties of an 8090 Al-Li alloy reinforced with 15 vol pet SiC particles were investigated, together with those of the unreinforced alloy processed following the same route. Two different heat treatments (naturally aged at ambient temperature and artificially aged at elevated temperature to the peak strength) were chosen because they lead to very different behaviors. Special emphasis was given to the analysis of the differences and similarities in the microstructure and in the deformation and failure mechanisms between the composite and the unreinforced alloy. It was found that the dispersion of the SiC particles restrained the formation of elongated grains during extrusion and inhibited the precipitation of Al3Li at ambient temperature. The deformation processes in the peak-aged materials were controlled by the S' precipitates, which acted as barriers for dislocation motion and homogenized the slip. Homogeneous slip was also observed in the naturally aged composite, but not in the unreinforced alloy, where plastic deformation was concentrated in slip bands. The most notorious differences between the alloy and the composite were found in the fracture mechanisms. The naturally aged unreinforced alloy failed by transgranular shear, while the failure of the peak-aged alloy was induced by grain-boundary fracture. The fracture of the composite in both tempers was, however, precipitated by the progressive fracture of the SiC reinforcements during deformation, which led to the early failure at the onset of plastic instability.

  3. Deformation Behavior of Laser Welds in High Temperature Oxidation Resistant Fe-Cr-Al Alloys for Fuel Cladding Applications

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G [ORNL; Gussev, Maxim N [ORNL; Yamamoto, Yukinori [ORNL; Snead, Lance Lewis [ORNL

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al in weight percent with a minor addition of yttrium using laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds has been carried out to determine the performance of welds as a function of alloy composition. Laser welding resulted in a defect free weld devoid of cracking or inclusions for all alloys studied. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. No significant correlation was found between the deformation behavior/mechanical performance of welds and the level of Cr or Al in the alloy ranges studied.

  4. Effects of molybdenum on microstructural evolution and mechanical properties in Zr–Nb alloys as nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    The Zr–Nb alloys were modified by doping of Mo as a minor alloying element to seek for the nuclear fuel cladding materials with better characteristics. The effects of Mo on microstructural evolution and mechanical properties in Zr–Nb alloys were systematically investigated and elucidated. Results showed that the martensitic microstructure, a mixture of lath martensites and lens martensites with internal twins, was observed in the alloys quenched from β-phase. Width of the lath martensite reduced with the increasing Mo concentration, and the volume fraction of lens martensite increased with increase in the Mo concentration. After final annealing, a new kind of precipitate, namely β-(Nb, Mo, Zr), was identified in the Mo-containing alloys. It was also found that Mo reduced the growth of the precipitates but increased their number density. Furthermore, Mo addition retarded the recrystallization process strongly and reduced the grain size significantly. In terms of the mechanical properties, Mo addition enhanced the yield strength and the ultimate tensile strength at room temperature, however decreased the ductility. The grain size strengthening was presumed as the greatest contributor in this system. (author)

  5. High temperature deformation of zircaloy-4 and Zr-Sn-Fe-Nb alloy cladding tubes

    International Nuclear Information System (INIS)

    In order to investigate the effect of dynamic strain aging on the high temperature deformation behavior of Zircaloy-4 and Zr-Sn-Fe-Nb nuclear fuel claddings, high temperature mechanical testing was carried out over the temperature range 298 ∼798K. The strengths of Zr-Sn-Fe-Nb claddings were greater than those of Zircaloy-4 over the whole temperature range with the ductilities of Zr-Sn-Fe-Nb claddings slightly lower then those of Zircaloy-4. The plateau of the straight was observed in both Zircaloy-4 and Zr-Sn-Fe-Nb claddings although the plateau behavior was more pronounced in Zr-Sn-Fe- Nb claddings. The loss of the ductility associated with dynamic strain aging was observed in the same temperature range where the plateau was observed. SEM observation revealed that the fracture surfaces of both Zircaloy-4 and Zr-Sn-Fe-Nb claddings were ductile irrespective of strain rate and temperature. The predicted yield strength and elongation were in good agreement with the experimental data, supporting that the yield stress plateau and the ductility loss are associated with dynamic strain aging

  6. Super ODS steels R and D for fuel cladding of next generation nuclear systems. 1) Introduction and alloy design

    International Nuclear Information System (INIS)

    Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, 'Super ODS steel' that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of 'Super ODS steel' R and D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of 'Super ODS steel'. Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700degC in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. 'Super ODS steel' is promising for the fuel cladding material of next generation nuclear systems, and the R and D is now ready to proceed to the next stage of empirical verification. (author)

  7. The design of cobalt-free, nickel-based alloy powder (Ni-3) used for sealing surfaces of nuclear power valves and its structure of laser cladding coating

    International Nuclear Information System (INIS)

    Research highlights: → The Ni-3 Co-free alloy coating prepared by laser welding. → Ni-3 alloy has excellent combination with stainless steel base. → Ni-3 alloy containing those strengthening phases could have excellent wear resistance and anti-oxidation ability at high temperature. - Abstract: To meet the demand of cobalt-free for the cladding coating materials used on sealing surface of nuclear power valves, a new Co-free, Ni-Cr based alloy powder (Ni-3) has been developed. It has been successfully coated on the surface of stainless steel as the strengthening layer. The XRD result reveals that the primary phase of cladding coating is Ni-based solid solution, and the carbides M7C3 and M23C6 as well as several A3B types of γ' strengthening phases. It indicates that the alloy possesses the high wear resistance, good corrosion resistance and high temperature tolerance. The test results suggest that the micro-hardness of Ni-3 corresponds to that of alloy Stellite 6 which containing cobalt and currently used as material for nuclear power valves. Hence, the developed Ni-3 alloy powder can be the hopeful candidate material for Co-free cladding material used on the surface of nuclear power valves; it can reduce the nuclear pollution and save the expensive metals.

  8. The design of cobalt-free, nickel-based alloy powder (Ni-3) used for sealing surfaces of nuclear power valves and its structure of laser cladding coating

    Energy Technology Data Exchange (ETDEWEB)

    Fu Geyan, E-mail: fugeyan@suda.edu.c [School of Mechanical and Electric Engineering, Soochow University, Suzhou 215021 (China); Liu Shuang [School of Mechanical and Electric Engineering, Soochow University, Suzhou 215021 (China); Fan Jiwei [School of Materials Science and Chemical Engineering, Zhongyuan University of Technology, Zhengzhou 450007 (China)

    2011-05-15

    Research highlights: The Ni-3 Co-free alloy coating prepared by laser welding. Ni-3 alloy has excellent combination with stainless steel base. Ni-3 alloy containing those strengthening phases could have excellent wear resistance and anti-oxidation ability at high temperature. - Abstract: To meet the demand of cobalt-free for the cladding coating materials used on sealing surface of nuclear power valves, a new Co-free, Ni-Cr based alloy powder (Ni-3) has been developed. It has been successfully coated on the surface of stainless steel as the strengthening layer. The XRD result reveals that the primary phase of cladding coating is Ni-based solid solution, and the carbides M{sub 7}C{sub 3} and M{sub 23}C{sub 6} as well as several A{sub 3}B types of {gamma}' strengthening phases. It indicates that the alloy possesses the high wear resistance, good corrosion resistance and high temperature tolerance. The test results suggest that the micro-hardness of Ni-3 corresponds to that of alloy Stellite 6 which containing cobalt and currently used as material for nuclear power valves. Hence, the developed Ni-3 alloy powder can be the hopeful candidate material for Co-free cladding material used on the surface of nuclear power valves; it can reduce the nuclear pollution and save the expensive metals.

  9. A comparative wear study on Al-Li and Al-Li/SiC composite

    Energy Technology Data Exchange (ETDEWEB)

    Okumus, S. Cem, E-mail: cokumus@sakarya.edu.tr; Karslioglu, Ramazan, E-mail: cokumus@sakarya.edu.tr; Akbulut, Hatem, E-mail: cokumus@sakarya.edu.tr [Sakarya University Engineering Faculty, Department of Metallurgical and Materials Engineering, Esentepe Campus, 54187, Sakarya (Turkey)

    2013-12-16

    Aluminum-lithium based unreinforced (Al-8090) alloy and Al-8090/SiCp/17 vol.% metal matrix composite produced by extrusion after spray co-deposition. A dry ball-on disk wear test was carried out for both alloy and composite. The tests were performed against an Al{sub 2}O{sub 3} ball, 10 mm in diameter, at room temperature and in laboratory air conditions with a relative humidity of 40-60%. Sliding speed was chosen as 1.0 ms{sup −1} and normal loads of 1.0, 3.0 and 5.0 N were employed at a constant sliding distance of 1000 m. The wear damage on the specimens was evaluated via measurement of wear depth and diameter. Microstructural and wear characterization was carried out via scanning electron microscopy (SEM). The results showed that wear loss of the Al-8090/SiC composite was less than that of the Al-8090 matrix alloy. Plastic deformation observed on the wear surface of the composite and the matrix alloy, and the higher the applied load the greater the plastic deformation. Scanning electron microscopy examinations of wear tracks also reveal that delamination fracture was the dominant wear mechanism during the wear progression. Friction coefficient was maximum at the low applied load in the case of the Al-8090/SiC composite while a gradual increase was observed with applied load for the matrix alloy.

  10. Deformation behavior of laser welds in high temperature oxidation resistant Fe–Cr–Al alloys for fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G., E-mail: fieldkg@ornl.gov; Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Yamamoto, Yukinori, E-mail: yamamotoy@ornl.gov; Snead, Lance L., E-mail: sneadll@ornl.gov

    2014-11-15

    Ferritic-structured Fe–Cr–Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe–(13–17.5)Cr–(3–4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  11. Inconel alloy 625 clad steel for application in wet scrubber systems

    International Nuclear Information System (INIS)

    Test panels from INCONEL 625 clad plate were successfully installed in two wet flue gas scrubber systems. In one system INCONEL 625 clad plate was located in the roof section of the absorber just ahead of the outlet ducting. The test plates, including weld seams, showed no signs to corrosion after six months of exposure. In the other scrubber test plates located in the outlet duct of an I.D. fan house, in the stack lining, and in the absorber quench area were unattacked after nine months

  12. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  13. Effect of tool profile and fatigue loading on the local hardness around scratches in clad and unclad aluminium alloy 2024

    International Nuclear Information System (INIS)

    Nanoindentation has been used to study the hardness changes produced by scratching of aluminium alloy AA2024, with and without a clad layer of pure aluminium. The hardness was mapped around scratches made with diamond tools of different profiles. One tool produced significant plastic damage with associated hardening at the scratch root, whilst the other produced a 'cleaner' cut with no hardening. The different behaviours are attributed to whether the tool makes the scratch by a 'cutting' or a 'ploughing' mechanism. The degree of plastic damage around the scratches has been correlated with peak broadening data obtained using synchrotron X-ray diffraction. There was no change observed in the local hardness around the scratch with fatigue loading.

  14. Microstructure and high temperature oxidation resistance of Ti-Ni gradient coating on TA2 titanium alloy fabricated by laser cladding

    Science.gov (United States)

    Liu, Fencheng; Mao, Yuqing; Lin, Xin; Zhou, Baosheng; Qian, Tao

    2016-09-01

    To improve the high temperature oxidation resistance of TA2 titanium alloy, a gradient Ni-Ti coating was laser cladded on the surface of the TA2 titanium alloy substrate, and the microstructure and oxidation behavior of the laser cladded coating were investigated experimentally. The gradient coating with a thickness of about 420-490 μm contains two different layers, e.g. a bright layer with coarse equiaxed grain and a dark layer with fine and columnar dendrites, and a transition layer with a thickness of about 10 μm exists between the substrate and the cladded coating. NiTi, NiTi2 and Ni3Ti intermetallic compounds are the main constructive phases of the laser cladded coating. The appearance of these phases enhances the microhardness, and the dense structure of the coating improves its oxidation resistance. The solidification procedure of the gradient coating is analyzed and different kinds of solidification processes occur due to the heat dissipation during the laser cladding process.

  15. Effects of Heat Treatment on Microstructure and Hardness of Laser Clad NiWCRE Alloy Layer

    Institute of Scientific and Technical Information of China (English)

    LIU Su-qin; HUANG Jin-liang; WANG Shun-xing; DONG Qi-ming

    2004-01-01

    The effects of heat treatment on microstructure and hardness of laser surface-clad Ni21+20%WC+0.5%CeO2 on the heat-resistant cast iron were investigated by means of X-ray diffraction(XRD), transmission electron microscope(TEM)and microhardness test. The experimental results showed that heat-treating at 500℃ has no effect on microstructure and hardness of the layers. Although the phase composition of the layers heat-treated at 700℃ and 800℃ remain unchanged,more Ni3B and Ni4B3 phases are precipitated on the matrix of the cladding layer, the metastable phase-M7C3 is transformed into steady phase-M23C6, and the precipitated phases coarsened.

  16. Effects of Heat Treatment on Microstructure and Hardness of Laser Clad NiWCRE Alloy Layer

    Institute of Scientific and Technical Information of China (English)

    LIUSu-qin; HUANGJin-liang; WANGShun-xing; DONGQi-ming

    2004-01-01

    The ettects of heat treatment on microstructure and hardness ot laser surface-clad Ni21+20%WC+0.5%CeO2 on the heat-resistant cast iron were investigated by means of X-ray diffraction(XRD), transmission electron microscope(TEM) and microhardness test. The experimental results showed that heat-treating at 500℃ has no effect on microstructure and hardness of the layers. Although the phase composition of the layers heat-treated at 700℃ and 800℃ remain unchanged, more Ni3B and Ni4B3 phases are precipitated on the matrix of the cladding layer, the metastable phase-M7C3 is transformed into steady phase-M23C6, and the precipitated phases coarsened.

  17. Microstructure and Tribological Properties of In Situ Synthesized TiN Reinforced Ni/Ti Alloy Clad Layer Prepared by Plasma Cladding Technique

    Science.gov (United States)

    Jin, Guo; Li, Yang; Cui, Huawei; Cui, Xiufang; Cai, Zhaobing

    2016-06-01

    A Ni/Ti composite coating enhanced by an in situ synthesized TiN phase was fabricated on FV520B steel by plasma cladding technology. The in situ formation of the TiN phase was confirmed by XRD, SEM, and TEM. The cladding layer consisted of three regions on going from the top to the bottom, namely, columnar grain regions, columnar dendritic regions, and fine grain regions. The cladding layer was composed of Ni3Ti, TiN, (Fe, Ni), and Ti phases. The dendritic and columnar regions were mainly composed of the Ni3Ti and (Fe, Ni) phases. The Ti phase was observed at the branches of dendrite crystals and columnar grains. The volume fraction of the TiN phase in the cladding layer was about 3.2%. The maximum micro-hardness value of the in situ formed coating (760 HV0.2) was higher than that of the pure coating (537 HV0.2). The cladding layer had a small amount of scratch and wear debris when a load of 20 N was used. As the test load increased, the wear debris in the cladding layer also increased and the massive furrows were not observed.

  18. Investigation of grain boundary chemistry in Al-Li 2195 welds using Auger electron spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, J.H. [National Aeronautics and Space Administration, Huntsville, AL (United States). George C. Marshall Space Flight Center

    1996-05-01

    Al-Li alloy 2195 is a low-density material with high fracture toughness that is particularly well-suited for aerospace systems. It will replace Al-Cu alloy 2219 in the Super Light Weight Tank (SLWT), a modified version of the external tank being developed for the Space Shuttle to support Space Station deployment. Recent efforts have focused on joining 2195 with variable polarity plasma arc welding, as well as repairing 2195 welds with tungsten inert gas techniques. During this study, Auger electron spectroscopy (AES) was used to examine grain boundary chemistry in 2195 welds. Results indicated that weld integrity depends on whether (and how much) the grain boundaries are covered with thin films comprised of a mixture of discontinuous Al{sub 2}O{sub 3} in Al (Al/Al{sub 2}O{sub 3}), which form during weld solidification. O was probably introduced as a contaminant in the shielding gases, occurring at low levels considered negligible for Al alloys that do not contain Li. However, oxidation kinetics in 2195 are increased by Li enrichment of small quantities of Al{sub 2}O{sub 3}, further enhancing thin film formation at the grain boundaries. Al{sub 2}O{sub 3} can ultimately occupy sufficient grain boundary area to degrade the material`s mechanical properties, producing negative effects that are compounded by the cumulative heat input of multi-pass repair welding. (orig.)

  19. Laser-clad Ni70Al20Cr7Hf3 alloys with extended solid solution of Hf: Part I. Microstructure evolution

    Science.gov (United States)

    Sircar, S.; Ribaudo, C.; Mazumder, J.

    1989-11-01

    Coatings for superalloys for extended service in atmospheres at high temperature are of great interest at present. The addition of reactive elements (RE’s) such as Hf to these coatings has a pronounced effect on their high-temperature oxidation resistance. A laser-cladding technique was used to produce Ni-Al-Cr-Hf alloys with an extended solid solution of Hf in a nearstoichiometric Ni3Al matrix. A 10 kW CO2 laser with mixed powder feed was used for the cladding process. Scanning electron microscope (SEM), transmission electron microscope (TEM), and scanning transmission electron microscope (STEM) were employed for studies of microstructural evolution of alloys produced during the laser-cladding process. Microstructural studies reveal the formation of dendrites with a solid solubility of about 11 to 14 wt pct Hf and also a eutectic structure. Convergent-beam techniques and X-ray spectroscopy have been applied to characterize the phases formed during the cladding process.

  20. Modelling of behaviour of 37 fuel rod assembly with Zr1%Nb-alloy simulators cladding under loss-of-coolant accident conditions on PARAMETR-M facility

    International Nuclear Information System (INIS)

    The experiment described in this report involves the implementation of conditions complying with the second stage of LOCA accident, for representative group of WWER-1000 fuel rods with relative heat generation rate in the range of 1.2-1.4 from the average one: maximal cladding temperature up to 9000C. The testing of experimental fuel rod assembly consisting of 37 fuel elements with Zr1%Nb-alloyed claddings has been made for representative group of heat-stressed fuel rods of the WWER-1000 type reactor on the electro heated PARAMETR-M facility under LOCA simulating conditions. The cladding rupture of fuel rods took place at the heating-up stage within the stated temperature interval 800-9000C. There were identified the basic cladding deformation and rupture parameters: temperature, pressure, axial distribution of hoop strain, and azimuthal distributions of radial deformation in rupture section. The experimental and calculated value of cross section blockage in the assembly under testing was 38%. The calculated values of cladding deformation and rupture parameters determined using RAPTA-5 Code agree well with experimental ones

  1. Development of aluminum (Al5083)-clad ternary Ag-In-Cd alloy for JSNS decoupled moderator

    Energy Technology Data Exchange (ETDEWEB)

    Teshigawara, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)]. E-mail: teshigawara.makoto@jaea.go.jp; Harada, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Saito, S. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Oikawa, K. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Maekawa, F. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Futakawa, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kikuchi, K. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kato, T. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Ikeda, Y. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Naoe, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Koyama, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Ooi, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Zherebtsov, S. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Kawai, M. [High Energy Accelerator Research Organization, 1-1, Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Narita-machi, Oarai-machi, Higashi ibaraki-gun, Ibaraki 311-1313 (Japan); Konashi, K. [International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Narita-machi, Oarai-machi, Higashi ibaraki-gun, Ibaraki 311-1313 (Japan)

    2006-09-15

    To develop Ag (silver)-In (indium)-Cd (cadmium) alloy decoupler, a method is needed to bond the decoupler between Al alloy (Al5083) and the ternary Ag-In-Cd alloy. We found that a better HIP condition was temperature, pressure and holding time at 803 K, 100 MPa and 10 min. for small test pieces ({phi}22 mm in dia. x 6 mm in height). Hardened layer due to the formation of AlAg{sub 2} was found in the bonding layer, however, the rupture strength of the bonding layer is more than 30 MPa, the calculated design stress. Bonding tests of a large size piece (200 x 200 x 30 mm{sup 3}), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength.

  2. Development of aluminum (Al5083)-clad ternary Ag In Cd alloy for JSNS decoupled moderator

    Science.gov (United States)

    Teshigawara, M.; Harada, M.; Saito, S.; Oikawa, K.; Maekawa, F.; Futakawa, M.; Kikuchi, K.; Kato, T.; Ikeda, Y.; Naoe, T.; Koyama, T.; Ooi, T.; Zherebtsov, S.; Kawai, M.; Kurishita, H.; Konashi, K.

    2006-09-01

    To develop Ag (silver)-In (indium)-Cd (cadmium) alloy decoupler, a method is needed to bond the decoupler between Al alloy (Al5083) and the ternary Ag-In-Cd alloy. We found that a better HIP condition was temperature, pressure and holding time at 803 K, 100 MPa and 10 min. for small test pieces ( ϕ22 mm in dia. × 6 mm in height). Hardened layer due to the formation of AlAg 2 was found in the bonding layer, however, the rupture strength of the bonding layer is more than 30 MPa, the calculated design stress. Bonding tests of a large size piece (200 × 200 × 30 mm 3), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength.

  3. On the dynamic phase transition for Nb-containing cladding alloys

    International Nuclear Information System (INIS)

    The licensing of new materials requires modifications to fuel performance codes that are extensively used by industry and safety authorities to verify compliance with the fuel safety criteria. Nevertheless, two types of fuel performance codes are generally being applied in the licensing process, corresponding to the normal operation and the design basis accident (DBA) conditions respectively. In order to simplify the code management by limiting the number of programs and in order to take advantage of the hardware improvements, one should generate a single fuel performance code that can cope with the different conditions. On one hand, extending the application range of a fuel performance code originally developed for steady-state conditions to accident conditions requires modifications to the basic equations in the thermalmechanical description of the fuel rod behaviour, stable numerical algorithms and a proper time-step control, in addition to the implementation of specific models dealing with the high temperature behaviour of cladding such as observed under loss of coolant (LOCA) conditions. On the other hand, for fuel performance codes developed to simulate some aspects of the nuclear fuel behaviour under accident conditions, such as TESPA, MFPR, or FRAPTRAN, either the thermo-mechanical behaviour of the fuel must be incorporated and/or the extension of models to normal operating conditions is necessary to consider burnup dependent phenomena such as thermal conductivity degradation, fission gas release and swelling as well as cladding corrosion. Such a posteriori modifications of the fuel performance code may entail difficulties in terms of convergence and calculation time. (orig.)

  4. The Role of X-Ray Diffraction for Analyzing Zr-Sn-Nb-Fe Alloys as Power Reactor Fuel Cladding

    International Nuclear Information System (INIS)

    Synthesis of Zr- 1%Nb- 1%Sn- 1%Fe alloy is undertaken in order to develop fuel cladding alloy at high burn-up. Powder specimens of Zr-Sn-Nb-Fe alloy were prepared and then formed into pellets with a dimension of 10 mm in height x 10 mm in diameter using a pressure of 1.2 ton/cm2. The 5 gram green pellets were then melted in an arc furnace crucible under argon atmosphere. The pressure in the furnace was set at 2 psi and the current was 50 A. Afterwards, the ingots were heated at a temperature of 1100 oC for 2 hours and subsequently quenched in water. The ingots then underwent annealing at temperatures of 400 oC, 500 oC, 600 oC, 700 oC, and 750 oC for 2 hours. The specimens were analyzed using X-ray diffraction in order to construct diffractograms. Results of the diffraction patterns were fitted with data from JCPDF (Joint Committee Powder Diffraction File) to determine the type of crystals in the elements or substances. The greater the crystallite dimension, the smaller the dislocation density. Agreeable results for hardening or strengthening were obtained at annealing temperatures of 500 oC and 700, whereas for softening or residual stress at 600 oC and 750 oC. The nucleation of the secondary phase precipitate (SPP) was favourable at annealing temperatures of 400 oC, 500 oC, and 700 oC. For Zr- 1%Nb- 1%Sn- 1%Fe alloy with annealing temperatures between 400 oC to 800 oC, precipitates of Fe2Nb, ZrSn2,FeSn, SnZr, NbSn2, Zr0.68Nb0.25Fe0.08, Fe2Nb0.4Zr0.6, Fe37Nb9Zr54, and ω-Zr were observed. Satisfactory precipitate stabilization was achieved at annealing temperature of 800 oC, growth of precipitates at temperature between 500 oC to 600 oC, and minimization of precipitate size at 700 oC. (author)

  5. The Role of X-Ray Diffraction for Analyzing Zr-Sn-Nb-Fe Alloys as Power Reactor Fuel Cladding

    Directory of Open Access Journals (Sweden)

    Sugondo

    2010-08-01

    Full Text Available Synthesis of Zr-1%Nb-1%Sn-1%Fe alloy is undertaken in order to develop fuel cladding alloy at high burn-up. Powder specimens of Zr-Sn-Nb-Fe alloy were prepared and then formed into pellets with a dimension of 10 mm in height 10 mm in diameter using a pressure of 1.2 ton/cm2. The 5 gram green pellets were then melted in an arc furnace crucible under argon atmosphere. The pressure in the furnace was set at 2 psi and the current was 50 A. Afterwards, the ingots were heated at a temperature of 1100°C for 2 hours and subsequently quenched in water. The ingots then underwent annealing at temperatures of 400°C, 500°C, 600°C, 700°C, and 750°C for 2 hours. The specimens were analyzed using X-ray diffraction in order to construct diffractograms. Results of the diffraction patterns were fitted with data from JCPDF (Joint Committee Powder Diffraction File to determine the type of crystals in the elements or substances. The greater the crystallite dimension, the smaller the dislocation density. Agreeable results for hardening or strengthening were obtained at annealing temperatures of 500°C and 700, whereas for softening or residual stress at 600°C and 750°C. The nucleation of the secondary phase precipitate (SPP was favourable at annealing temperatures of 400°C, 500°C, and 700°C. For Zr-1%Nb-1%Sn-1%Fe alloy with annealing temperatures between 400°C to 800°C, precipitates of Fe2Nb, ZrSn2,FeSn, SnZr, NbSn2, Zr0.68Nb0.25Fe0.08, Fe2Nb0.4Zr0.6, Fe37Nb9Zr54, and ω-Zr were observed. Satisfactory precipitate stabilization was achieved at annealing temperature of 800°C, growth of precipitates at temperature between 500°C to 600°C, and minimization of precipitate size at 700°C.

  6. FY-13 FCRD Milestone M3FT-13OR0202311 Weldability of ORNL Accident Tolerant Fuel Cladding Model Alloys For Thin Walled Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G [ORNL; Gussev, Maxim N [ORNL; Yamamoto, Yukinori [ORNL

    2013-07-01

    Ferritic FeCrAl-based alloys show increased oxidation resistance for accident tolerant applications as fuel cladding. This study focuses on investigating the weldability of three model FeCrAl alloys with varying alloy compositions using laser-welding techniques. A detailed study of the mechanical properties of bead-on-plate welds was used to determine the quality of welds as a function of alloy composition. Laser welding resulted in defect free welds devoid of cracking or inclusions. Initial results indicate a reduction in the yield strength of weldments compared to the base material due to distinct changes in the microstructure within the fusion zone. Although a loss of yield strength was observed, there was no significant difference in the magnitude of the tensile property changes with varying Cr or Al content. Also, there was no evidence of embrittlement; the material in the fusion zones demonstrated ductile behavior with high local ductility.

  7. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

  8. Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

    International Nuclear Information System (INIS)

    The technical bases for specification of limits to environmental exposure conditions to avoid excessive degradation are being developed for storage criteria for dry storage of highly-enriched, aluminum-clad spent nuclear fuels owned by the US Department of Energy. Corrosion of the aluminum cladding is a limiting degradation mechanism (occurs at lowest temperature) for aluminum exposed to an environment containing water vapor. Attendant radiation fields of the fuels can lead to production of nitric acid in the presence of air and water vapor and would exacerbate the corrosion of aluminum by lowering the pH of the water solution. Laboratory-scale specimens are being exposed to various conditions inside an autoclave facility to measure the corrosion of the fuel matrix and cladding materials through weight change measurements and metallurgical analysis. In addition, electrochemical corrosion tests are being performed to supplement the autoclave testing by measuring differences in the general corrosion and pitting corrosion behavior of the aluminum cladding alloys and the aluminum-uranium fuel materials in water solutions

  9. Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor

    International Nuclear Information System (INIS)

    For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

  10. X-ray microtomography of fatigue crack closure as a function of applied load in Al-Li 2090 T8E41 samples

    Energy Technology Data Exchange (ETDEWEB)

    Morano, R.; Stock, S.R.; Davis, G.R.; Elliott, J.C.

    2000-07-01

    Crack closure is held to be responsible for very low fatigue crack growth rates in many alloys such as Al-Li 2090 T8E41, and early crack face contact during unloading or prolonged contact during loading seems to reduce the driving force for crack extension. High resolution x-ray computed tomography (i.e., microtomography) allows one to image the entire volumes of samples and to quantify opening as a function of applied load over the entire crack surface. Crack closure results are reported for a fatigue crack grown under load ratio R = 0.1 in a compact tension sample of Al-Li 2090 T8E41; the crack was free to choose its path unconstrained by side-grooves which are normally used to suppress crack deflection. The inter-relationship between crack path, crack face contact and applied load level are discussed.

  11. Cool-down induced hydride reorientation of hydrogen-charged Zirconium alloy cladding tubes

    International Nuclear Information System (INIS)

    250 and 500ppm hydrogen-charged Zirconium alloy tubes were employed to investigate hydride reorientation behaviors when they were cool down from 400 to 300, 200degC and room temperature with various cooling rates of 0.3, 2.0, 4.0, 7.0 and 15.0degC/min under a tensile hoop stress of 150MPa. These cool-down tests indicate that the slower cooling rate and the lower terminal cool-down temperature produced the more hydrides precipitated along with the larger fraction and the longer length of radial hydrides. These phenomena may be explained by terminal solid solubility of hydrogen for dissolution and precipitation and cooling rate-dependent hydride nucleation and growth rates. On the other hand, a dramatic decrease of ultimate tensile strength and plastic strain of the cool-down tested specimens may be explained by the amount of the radial hydrides precipitated during the cool-down process. (author)

  12. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    Science.gov (United States)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with 50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

  13. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    International Nuclear Information System (INIS)

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  14. Results of U-xMo (x=7, 10, 12 wt.%) Alloy versus Al-6061 Cladding Diffusion Couple Experiments Performed at 500, 550 and 600 Degrees C

    International Nuclear Information System (INIS)

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been developing low enrichment fuel systems encased in Al 6061 for use in research and test reactors. U-Mo alloys in contact with Al and Al alloys can undergo diffusional interactions that can result in the development of interdiffusion zones with complex fine-grained microstructures composed of multiple phases. A monolithic fuel currently being developed by the RERTR program has local regions where the U-Mo fuel plate is in contact with the Al 6061 cladding and, as a result, the program finds information about interdiffusion zone development at high temperatures of interest. In this study, the microstructural development of diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo, and U-12wt.%Mo vs. Al 6061 (or 6061 aluminum) cladding, annealed at 500, 550, 600 degrees C for 1, 5, 20, 24, or 132 hours, was analyzed by backscatter electron microscopy and x-ray energy dispersive spectroscopy on a scanning electron microscope. Concentration profiles were determined by standardized wavelength dispersive spectroscopy and standardless x-ray energy dispersive spectroscopy. The results of this work shows that the presence of surface layers at the U-Mo/Al 6061 interface can dramatically impact the overall interdiffusion behavior in terms of rate of interaction and uniformity of the developed interdiffusion zones. It further reveals that relatively uniform interaction layers with higher Si concentrations can develop in U-Mo/Al 6061 couples annealed at shorter times and that longer times at temperature result in the development of more non-uniform interaction layers with more areas that are enriched in Al. At longer annealing times and relatively high temperatures, U-Mo/Al 6061 couples can exhibit more interaction compared to U-Mo/pure Al couples. The minor alloying constituents in Al 6061 cladding can result in the development of many complex phases in the interaction layer of U-Mo/Al-6061 cladding

  15. Results of U-xMo (x=7, 10, 12 wt.%) Alloy versus Al-6061 Cladding Diffusion Couple Experiments Performed at 500, 550 and 600 Degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Yongho Sohn

    2013-04-01

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been developing low enrichment fuel systems encased in Al 6061 for use in research and test reactors. U–Mo alloys in contact with Al and Al alloys can undergo diffusional interactions that can result in the development of interdiffusion zones with complex fine-grained microstructures composed of multiple phases. A monolithic fuel currently being developed by the RERTR program has local regions where the U–Mo fuel plate is in contact with the Al 6061 cladding and, as a result, the program finds information about interdiffusion zone development at high temperatures of interest. In this study, the microstructural development of diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo, and U-12wt.%Mo vs. Al 6061 (or 6061 aluminum) cladding, annealed at 500, 550, 600 degrees C for 1, 5, 20, 24, or 132 hours, was analyzed by backscatter electron microscopy and x-ray energy dispersive spectroscopy on a scanning electron microscope. Concentration profiles were determined by standardized wavelength dispersive spectroscopy and standardless x-ray energy dispersive spectroscopy. The results of this work shows that the presence of surface layers at the U–Mo/Al 6061 interface can dramatically impact the overall interdiffusion behavior in terms of rate of interaction and uniformity of the developed interdiffusion zones. It further reveals that relatively uniform interaction layers with higher Si concentrations can develop in U–Mo/Al 6061 couples annealed at shorter times and that longer times at temperature result in the development of more non-uniform interaction layers with more areas that are enriched in Al. At longer annealing times and relatively high temperatures, U–Mo/Al 6061 couples can exhibit more interaction compared to U–Mo/pure Al couples. The minor alloying constituents in Al 6061 cladding can result in the development of many complex phases in the interaction layer of U

  16. 快堆先进包壳材料ODS合金发展研究%R &D on advanced cladding materials ODS alloys for fast reactor

    Institute of Scientific and Technical Information of China (English)

    崔超; 黄晨; 苏喜平; 宿彦京

    2011-01-01

    Fast reactor advanced cladding materials ODS alloys (Oxide Dispersion Strengthened steel) have excellent irradiation swelling resistance and stable mechanical properties at elevated temperature, which is chosen as the candidate cladding material of high burnup fuel for fast reactor. This paper generally introduces the progress of R&D on ODS alloys, including the processing technology of ODS alloys, mechanical properties, compatibility with sodium, irradiation performance and so on.%快堆先进包壳材料ODS合金(Oxide Dispersion Strengthened Steel)具有优异的抗辐照肿胀性能和高温力学性能,是高性能快堆燃料元件包壳管的主要候选材料.本文概括介绍了ODS合金的研究进展,包括ODS合金的制备方法、力学性能、与钠相容性以及辐照性能等.

  17. Microstructure and Wear Resistance of in situ NbC Particles Reinforced Ni-based Alloy Composite Coating by Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    DONG Gang; YAN Biao; DENG Qilin; YU Ting

    2012-01-01

    The in situ synthesized NbC particles reinforced Ni-based alloy composite coating was produced by laser cladding a precursor mixture of Ni-based alloy powder,graphite and niobium powders on a steel substrate.The microstructure,phase composition and wear property of the composite coating were investigated by means of scanning electron microscopy (SEM),X-ray diffraction (XRD) and dry sliding wear test.The experiment results show that the composite coating is homogeneous and free from cracks,and about 0.8 mm thick.The microstructure of the composite coating is mainly composed of NbC particles,CrB type chromium borides,γ-Ni primary dendrites,and interdendritic eutectics.CrB phases often nucleate and grow on the surface of NbC particles or in their close vicinity.NbC particles are formed via in situ reaction between niobium and graphite in the molten pool during the laser cladding process and they are commonly precipitated in three kinds of morphologies,such as quadrangle,cluster,and flower-like shape.Compared with the pure Nibased alloy coating,the microhardness of the composite coating is increased about 38%,giving a high average hardness of HV0.21000,and the wear rate of the composite coating is decreased by about 32%,respectively.These are attributed to the presence of in situ synthesized NbC particles and their well distribution in the coating.

  18. A comparative study on the high temperature corrosion of TP347H stainless steel, C22 alloy and laser-cladding C22 coating in molten chloride salts

    International Nuclear Information System (INIS)

    Highlights: • Two KCl and NaCl mixtures simulated molten salt corrosion of biomasses combustion. • The corrosivity of forestry and agricultural biomasses was comparatively studied. • Corrosion of TP347H, C22 alloy and C22 coating was carried out at 450–750 °C. • Laser-cladding C22 coating exhibited least performance degradation. • Microstructures, compositions and corrosion mechanisms were strongly interrelated. - Abstract: Isothermal corrosion of TP347H (A1), C22 alloy (A2) and laser-cladding C22 coating (A3) was evaluated by mass loss measurements in molten alkali chloride salts at 450–750 °C. Corrosion mechanisms were characterised by scanning electron microscopy, optical microscopy and X-ray diffraction. A3 exhibited superior corrosion resistance, followed by A2, which results from alloying elements, refined microstructure and Cr–O (CrOx), Co(Fe, Cr)2O4 in the corrosion scale. Severe intergranular corrosion caused failure of A1, slight intergranular corrosion happened in A2 but none in A3. Fe-rich oxides were main products of A1 while NiO of A2 and A3 with Cr2O3 and Mo-containing compositions

  19. Explosive Cladding of Titanium and Aluminium Alloys on the Example of Ti6Al4V-AA2519 Joints / Wybuchowe Platerowanie Stopów Tytanu I Aluminium Na Przykładzie Połączenia Ti6Al4V-AA2519

    Directory of Open Access Journals (Sweden)

    Gałka A.

    2015-12-01

    Full Text Available Explosive cladding is currently one of the basic technologies of joining metals and their alloys. It enables manufacturing of the widest range of joints and in many cases there is no alternative solution. An example of such materials are clads that include light metals such as titanium and aluminum. ach new material combination requires an appropriate adaptation of the technology by choosing adequate explosives and tuning other cladding parameters. Technology enabling explosive cladding of Ti6Al4V titanium alloy and aluminum AA2519 was developed. The clads were tested by means of destructive and nondestructive testing, analyzing integrity, strength and quality of the obtained joint.

  20. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... of materials, the structural features and the construction details of building systems in selected architectural works. With a particular focus at heavy constructions made of solid wood and masonry, and light weight constructions made of wooden frame structures and steel profiles, it is the intention...... tightness in constructions. At the same time a need for longevity and effortless maintenance have lead to contemporary architectural structures, where the exterior walls and the building envelope most often are made of several layers of advanced materials and separate building elements. In most contemporary...

  1. Modification of tribology and high-temperature behavior of Ti 48Al 2Cr 2Nb intermetallic alloy by laser cladding

    Science.gov (United States)

    Liu, Xiu-Bo; Wang, Hua-Ming

    2006-06-01

    In order to improve the tribology and high-temperature oxidation properties of the Ti-48Al-2Cr-2Nb intermetallic alloy simultaneously, mixed NiCr-Cr 3C 2 precursor powders had been investigated for laser cladding treatment to modify wear and high-temperature oxidation resistance of the material. The alloy samples were pre-placed with NiCr-80, 50 and 20%Cr 3C 2 (wt.%), respectively, and laser treated at the same parameters, i.e., laser output power 2.8 kW, beam scanning speed 2.0 mm/s, beam dimension 1 mm × 18 mm. The treated samples underwent tests of microhardness, wear and high-temperature oxidation. The results showed that laser cladding with different constitution of mixed precursor NiCr-Cr 3C 2 powders improved surface hardness in all cases. Laser cladding with NiCr-50%Cr 3C 2 resulted in the best modification of tribology and high-temperature oxidation behavior. X-ray diffraction (XRD), optical microscope (OM), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) analyses indicated that the formation of reinforced Cr 7C 3, TiC and both continuous and dense Al 2O 3, Cr 2O 3 oxide scales were supposed to be responsible for the modification of the relevant properties. As a result, the present work had laid beneficial surface engineering foundation for TiAl alloy applied as future light weight and high-temperature structural candidate materials.

  2. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Lee, Changhee, E-mail: chlee@hanyang.ac.kr [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Woo, WanChuck [Neutron Science Division, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Park, Sunhong [Research Institute of Industrial Science & Technology, Hyo-ja-dong, Po-Hang, Kyoung-buk, San 32 (Korea, Republic of)

    2015-08-01

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  3. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    International Nuclear Information System (INIS)

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  4. Non-destructive Residual Stress Analysis Around The Weld-Joint of Fuel Cladding Materials of ZrNbMoGe Alloys

    OpenAIRE

    Parikin; Bandriyana; I. Wahyono; A.H. Ismoyo

    2003-01-01

    The residual stress measurements around weld-joint of ZrNbMoGe alloy have been carried out by using X-ray diffraction technique in PTBIN-BATAN. The research was performed to investigate the structure of a cladding material with high temperature corrosion resistance and good weldability. The equivalent composition of the specimens (in %wt.) was 97.5%Zr1%Nb1%Mo½%Ge. Welding was carried out by using TIG (tungsten inert gas) technique that completed butt-joint with a current 20 amperes. Three reg...

  5. Aluminium-lithium alloys with scandium

    International Nuclear Information System (INIS)

    The influence of scandium on phase composition and properties of Al-Li alloys is considered. It is shown that the alloying with scandium increases strength properties, improves the weldability and affects the character and the velocity of decomposition of a supersaturated solid solution. The best compromise between strength and plastic properties is stated to be provided by combined introduction of Sc and Zr to Al-Li alloys

  6. Effect of yttrium additions on the elevated-temperature tensile properties and hardness of an advanced iron-nickel-chromium LMFBR cladding and duct alloy

    Energy Technology Data Exchange (ETDEWEB)

    Song, M.H.

    1981-10-01

    The effect of the addition of yttrium on the elevated temperature tensile properties and hardness of an Fe-34% Ni-12% Cr candidate LMFBR cladding and duct alloy was investigated. Tensile tests were performed from room temperature to 800/sup 0/C in 100/sup 0/C steps at strain rates of 2.2 x 10/sup -3/ and 2.2 x 10/sup -4/ sec/sup -1/. Hardness tests were performed from room temperature to 850/sup 0/C in 50/sup 0/C steps. The addition of 0.1% yttrium decreased the yield stress and ultimate tensile stress in the test temperature range employed. Hardness also decreased over this test temperature range. In tensile tests, dynamic strain aging behavior occurred both for the undoped and doped alloy in the temperature range from 200 to 600/sup 0/C and 300 to 600/sup 0/C for the lower and higher strain rate, respectively.

  7. Effect of yttrium additions on the elevated-temperature tensile properties and hardness of an advanced iron-nickel-chromium LMFBR cladding and duct alloy

    International Nuclear Information System (INIS)

    The effect of the addition of yttrium on the elevated temperature tensile properties and hardness of an Fe-34% Ni-12% Cr candidate LMFBR cladding and duct alloy was investigated. Tensile tests were performed from room temperature to 8000C in 1000C steps at strain rates of 2.2 x 10-3 and 2.2 x 10-4 sec-1. Hardness tests were performed from room temperature to 8500C in 500C steps. The addition of 0.1% yttrium decreased the yield stress and ultimate tensile stress in the test temperature range employed. Hardness also decreased over this test temperature range. In tensile tests, dynamic strain aging behavior occurred both for the undoped and doped alloy in the temperature range from 200 to 6000C and 300 to 6000C for the lower and higher strain rate, respectively

  8. Effect of Mo and nano-Nd2O3 on the microstructure and wear resistance of laser cladding Ni-based alloy coatings

    Science.gov (United States)

    Ding, Lin; Hu, Shengsun; Quan, Xiumin; Shen, Junqi

    2016-04-01

    Three kinds of coatings were successfully prepared on Q235 steel by laser cladding technique through adulterating with Mo and nano-Nd2O3 into Ni-based alloys. The effect of Mo and nano-Nd2O3 on the microstructure and properties of Ni-based coatings was investigated systematically by means of optical microscopy, X-ray diffraction, scanning electron microscopy, energy-dispersive spectroscopy, and microhardness testing and wear testing. The results indicated a certain amount of fine grains and polygonal equiaxed grains synthesized after adding Mo and nano-Nd2O3. Both the microhardness and wear resistance of Ni-based coatings improved greatly with a moderate additional amount of Mo and nano-Nd2O3. The largest improvement in microhardness was 31.9 and 14.7 %, and the largest reduction in loss was 45.0 and 30.7 %, respectively, for 5.0 wt% Mo powders and 1.0 wt% nano-Nd2O3. The effect of Mo on microhardness and wear resistance of laser cladding Ni-based alloy coatings is greater than the effect of nano-Nd2O3.

  9. Nd:YAG laser cladding of Co-Cr-Mo alloy on γ-TiAl substrate

    Science.gov (United States)

    Barekat, Masoud; Shoja Razavi, Reza; Ghasemi, Ali

    2016-06-01

    In this work, Co-Cr-Mo powder is used to form laser clads on a γ-TiAl substrate. The single-track geometrical characteristics such as width, height, penetration depth, dilution and wetting angle play the important role to control the characteristics of laser clad coatings formed by overlap of individual tracks. This paper is investigated the relations between the main coaxial laser cladding parameters (laser power P, laser beam scanning speed S and powder feeding rate F) and geometrical characteristics of single tracks by linear regression analysis. The results show that the clad height, H, depends linearly on the FS-5/4 parameter with the laser power having a minimal effect. Similarly, the cladding width W is controlled by the PS-2/3 parameter. The penetration depth b and dilution, D are proportional to P2S-1/4F-1/4 and P2/3S1/2F-1/2 respectively and wetting angle is controlled by the P1/4S1/2F-1/2 parameter. These empirical dependencies are observed with high values of the correlation coefficient (R>0.9). Finally, based on these relations, a laser cladd processing map was designed to use as a guideline for the selection of proper processing parameters for a required coating.

  10. Evolution of microstructure and properties in laser cladding of a Ni-Cr-B-Si hardfacing alloy

    NARCIS (Netherlands)

    Hemmati, I.; Ocelík, V.; De Hosson, J.T.M.

    2011-01-01

    Ni-Cr-B-Si coatings are used in many industrial applications in order to improve wear and/or corrosion properties. These coatings have traditionally been deposited by thermal spray techniques but the laser cladding process is also being increasingly employed to produce Ni-Cr-B-Si coatings with super

  11. Evaluation of corrosion and mechanical properties of Zr-Nb-Sn-Fe-X alloys for fuel claddings

    International Nuclear Information System (INIS)

    The corrosion resistance of Zr-Nb-Sn-Fe-X alloys were evaluated by the autoclave tests under the environments of 360 .deg. C water, 360 .deg. C LiOH 70 ppm solution and 400 .deg. C steam. The mechanical properties of those alloys were also investigated by tensile tests and creep tests. The corrosion resistance of the alloys in the water and the LiOH solution showed similar behavior, while they are superior to that of Zircaloy-4 in LiOH solution. The alloys, which have much in alloying content, showed better properties in tensile strength and creep resistance due to alloying effect. The final heat treatment of the alloys at 470 .deg. C and 520 .deg. C has little differences in corrosion behavior but much in mechanical strength and creep strength because the heat treatment at 470 .deg. C has more dislocation barrier than that at 520 .deg. C

  12. Microstructure and properties of the low-power-laser clad coatings on magnesium alloy with different amount of rare earth addition

    Science.gov (United States)

    Zhu, Rundong; Li, Zhiyong; Li, Xiaoxi; Sun, Qi

    2015-10-01

    Due to the low-melting-point and high evaporation rate of magnesium at elevated temperature, high power laser clad coating on magnesium always causes subsidence and deterioration in the surface. Low power laser can reduce the evaporation effect while brings problems such as decreased thickness, incomplete fusion and unsatisfied performance. Therefore, low power laser with selected parameters was used in our research work to obtain Al-Cu coatings with Y2O3 addition on AZ91D magnesium alloy. The addition of Y2O3 obviously increases thickness of the coating and improves the melting efficiency. Furthermore, the effect of Y2O3 addition on the microstructure of laser clad Al-Cu coatings was investigated by scanning electron microscopy. The energy-dispersive spectrometer (EDS) and X-ray diffractometer (XRD) were used to examine the elemental and phase compositions of the coatings. The properties were investigated by micro-hardness test, dry wear test and electrochemical corrosion. It was found that the addition of Y2O3 refined the microstructure. The micro-hardness, abrasion resistance and corrosion resistance of the coatings was greatly improved compared with the magnesium matrix, especially for the Al-Cu coating with Y2O3 addition.

  13. HIGH PURITY ALUMINIUM-LITHIUM MASTER ALLOY BY MOLTEN SALT ELECTROLYSIS

    OpenAIRE

    Watanabe, Y.; Toyoshima, M.; Itoh, K.

    1987-01-01

    The aim of this work is to develop the economical production process of the Al-Li master alloy free from metallic sodium, calcium and potassium. This master alloy can be used for aluminium-lithium alloys for structual materials of aircrafts, automobiles and robots. Moreover the Al-Li master alloy with lithium content of 18-20wt. % is applicable to the blanket of fusion reactors and the active mass of batteries. This Al-Li master alloy can be produced by means of LiCl-KCl molten salt electroly...

  14. Phase transformations in the Al-Li-Zr system

    International Nuclear Information System (INIS)

    Phase transformations in an Al-2.3Li-1.1Zr (wt%) alloy have been studied using electron microscopy techniques. The L12-ordered phase, or α', which is a stable precipitate at temperatures below the solidus, is described as Al3(Zr,Li) with varying Zr/Li ratios depending upon precipitation mechanism. Discontinuously precipitated filaments of α' have a Zr/Li atom ratio of about unity, whereas normal nucleation-and-growth α' has Zr/Li ratios of approximately 4:1. This compositional analysis was accomplished using transmission electron microscopy (TEM) image calculation techniques. The α' is typically perfectly coherent with the aluminum matrix and serves as a preferred nucleation site for Al3Li, or δ', precipitation when the alloy is aged at 1900C. The δ', also L12-ordered, nucleates in the variation which results in the minor sublattice being continuous across the α'/δ' interface. The fine α' distribution is extremely stable for extended periods at 4500C although it appears that a transformation to an equilibrium tetragonal phase is initiated during extended heat treatment

  15. Nuclear fuel element cladding

    International Nuclear Information System (INIS)

    Composite cladding for a nuclear fuel element containing fuel pellets is formed with a zirconium metal barrier layer bonded to the inside surface of a zirconium alloy tube. The composite tube is sized by a cold working tube reduction process and is heat treated after final reduction to provide complete recrystallization of the zirconium metal barrier layer and a fine-grained microstructure. The zirconium alloy tube is stress-relieved but is not fully recrystallized. The crystallographic structure of the zirconium metal barrier layer may be improved by compressive deformation such as shot-peening. (author)

  16. Surface Strengthening of Ta-C Compound Cladded for Tantalum Alloy%钽合金熔敷Ta-C化合物的表面强化

    Institute of Scientific and Technical Information of China (English)

    张小明; 白新房; 蔡小梅; 张于胜; 王峰; 王晖

    2012-01-01

    采用一种反应熔敷法,在钽合金表面制备一层较厚的碳化物强化层,进行钽合金的表面强化.以TaW12合金板材作为基板,将Ta粉和C粉混合,均匀涂覆在TaW12合金板材表面.以电弧为熔敷热源,在整个粉末涂覆面上进行逐行扫描,引发混合粉末与基板的反应熔融和固相扩散,获得表面硬化层.表面熔敷层完全致密,为熔融结晶状态.外表面呈周期性的波浪形,无裂纹,无气孔显现.熔敷层主要是Ta2C相与钽固溶体混合的共晶凝固组织,等轴状和条状的Ta2C相分布在Ta合金基体上,呈现一定的方向性,大多数近似垂直于板面.熔敷层与基体之间有一层由固态扩散产生的组织,大量针状Ta2C相在基体上交叉析出,形成网状结构.Ta2C相尺寸由外向内变小,逐步成为细小点状析出.熔敷层与过渡层间为固液界面,结合非常好,无裂纹和空洞存在.表面熔敷层平均维氏硬度为6690 MPa,达到了基体材料硬度的3倍.过渡层中硬度急剧下降,为熔敷层的一半.%A thick carbide layer has been prepared on a tantalum alloy surface for surface strengthening by means of reaction deposition. Ta and C powders were mixed and coated homogeneously on a TaW12 plate, and then layer-by-layer scanning was performed with electric arc as the heat producer to trigger solid phase diffusion and interreaction between mixed powders and the Ta matrix plate to obtain a surface hardening layer. The surface cladding layer is composed of molted crystalline and completely compact. Its outmost layer exhibits periodically wavy without cracks and pores. The cladding layer mainly is composed of the mixture of Ta2C and tantalum solid solution, and the equiaxed and streaky second-phases distribute on the matrix with a certain orientation, a majority of which approximately are vertical to plate face. There is a layer between the cladding layer and matrix, and a large number of needle Ta2C precipitation phases cross

  17. Evolution of Westinghouse fuel cladding

    International Nuclear Information System (INIS)

    As the nuclear power generating industry has matured, there is an increasing trend in core operating fuel duties. At the same time, refined requirements from regulators, e.g. in the areas of LOCA and RIA, must be fulfilled. This drives a continuing evolution of cladding materials, to provide more performance margin and support even higher fuel duty designs. Cladding performance, in particular with respect to in-reactor corrosion and hydrogen pickup, has improved dramatically since Zircaloy-2 and Zircaloy-4 were established in 1952 and 1960 respectively. For Westinghouse PWR cladding, the corrosion rate has decreased by more than one order of magnitude since; going from the original Zircaloy-4 to ZIRLO® and Optimized ZIRLO™ claddings. The next generation of Westinghouse PWR cladding, AXIOM™, shows further reduction of corrosion and hydrogen pickup, most notably at very high burnup, over 70 GWD/MTU. In Westinghouse BWR fuel, a carefully optimized variant of Zircaloy-2, LK3™ cladding, continues to demonstrate excellent performance under all operating conditions to date. In order to further reduce the hydrogen pickup, a new BWR cladding alloy, HiFi™, developed by NFI, is now being verified. Data indicate a reduction of the hydrogen absorption of around 50% with respect to Zircaloy-2. This paper describes the evolution of the different PWR and BWR cladding materials, providing details of their current experience base and post-irradiation examinations. (author)

  18. Development of advanced LWR fuel cladding - A study on the construction of phase diagram for multi-component Zr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seon Jin; Oh, Young Min; Jeong, Heung Sik [Hanyang University, Seoul (Korea)

    2000-03-01

    When the specimens were air-cooled at slow cooling rate, the width of - lath is increased as increasing the holding temperature of region. The addition of Sn, Nb, Fe and V resulted in the refining of the air-cooled microstructure while the addition of Sb and Mn led to the coarsening of the air-cooled microstructure. The transitions of the slipped to twinned martensite and the twinned martensite to basketweave structure were occurred in water-quenched Zr alloys as the Ms temperature of them varied with the amount and the kind of alloying elements. The addition of Nb in Zr alloys increased the recrystallizing temperature and, as a result, the recrystallization and the grain growth were suppressed. Although the recrystallization temperature gradually increased with increasing Sb content, and the suppression of recrystallization and grain growth were occurred, the effect of suppression was insignificant compared with Zr-0.8Sn-xNb alloys. In case that Sn was added into Zr-0.4Nb alloy, the solution limit seemed to generally decrease at the same temperature compared with Zr-Sn alloys and the regions of {alpha}, {alpha}+ppt., {alpha}+{beta}, {beta} were not much different from those of binary Zr-Sn alloys. In case that Nb was added into Zr-0.8Sn alloy, the eutectoid temperature showed a marked increase compared with the binary Zr-Nb alloys and the temperature of the regions of {alpha}, {alpha}+ppt., {alpha}+{beta}, {beta} increased as a result. 74 refs., 14 figs., 10 tabs. (Author)

  19. Experimental results on the interactions between hydrogen and zirconium claddings

    International Nuclear Information System (INIS)

    Experiments were performed with Zr1%Nb and Zircaloy-4 alloys to study the interaction between hydrogen and Zr containing cladding materials. Four main activities are summarised in the report: equilibrium solubility of hydrogen in cladding with oxygen content, escape of hydrogen during steam oxidation, escape of hydrogen during steam oxidation of cladding alloys with H-content, delaying effect of surface oxide layer on the hydrogen absorption from gas phase by the Zr alloys. (author)

  20. A study of TiB2/TiB gradient coating by laser cladding on titanium alloy

    Science.gov (United States)

    Lin, Yinghua; Lei, Yongping; Li, Xueqiao; Zhi, Xiaohui; Fu, Hanguang

    2016-07-01

    TiB2/TiB gradient coating has been fabricated by a laser cladding technique on the surface of a Ti-6Al-4V substrate using TiB2 powder as the cladding material. The microstructure and mechanical properties of the gradient coating were analyzed by SEM, EPMA, XRD, TEM and an instrument to measure hardness. With the increasing distance from the coating surface, the content of TiB2 particles gradually decreased, but the content of TiB short fibers gradually increased. Meanwhile, the micro-hardness and the elastic modulus of the TiB2/TiB coating showed a gradient decreasing trend, but the fracture toughness showed a gradient increasing trend. The fracture toughness of the TiB2/TiB coating between the center and the bottom was improved, primarily due to the debonding of TiB2 particles and the high fracture of TiB short fibers, and the fracture position of TiB short fiber can be moved to an adjacent position. However, the debonding of TiB2 particles was difficult to achieve at the surface of the TiB2/TiB coating.

  1. Recent developments of the aluminium-lithium system alloys for aircraft uses

    International Nuclear Information System (INIS)

    A brief review is made of the latest developments in the production of Aluminium-Lithium alloys. The necessity for new materials in the field of aeronautics has speeded up research on metallic and non-metallic materials. Lately, a good part of the research in the field of metallic components has been directed at Al-Li alloys. More recently, with the development of quaternary alloys Al-Li-X-X, the old problem of low toughness was overcome. The finality of this study is to cover the developments of the mentioned alloys, including the fundamentals of physical metallurgy of the complex system recently developed Al-Li-Cu-Mg. (author)

  2. Cold worked 15Cr15NiTiMoB alloys for cladding application in fast breeder reactors

    International Nuclear Information System (INIS)

    In this paper we will first give a quick overlook on the work achieved to develop CW15Cr15NiTiMoB for cladding applications in France and DeBeNe under national or cooperative programmes. Examples of the progress realised in the experimental data for core management design and safety conditions will be presented. The results achieved have emphasized that the initial batches of this material are at present qualified for at least 130 dpaNRT, the first improved generation should be acceptable to doses above 160 dpaNRT (dose already reached on a Phenix subassembly). The future common CEA-DeBe specification is expected to sustain the EFR target

  3. A study on wear resistance and microcrack of the Ti3Al/TiAl + TiC ceramic layer deposited by laser cladding on Ti-6Al-4V alloy

    International Nuclear Information System (INIS)

    Laser cladding of the Al + TiC alloy powder on Ti-6Al-4V alloy can form the Ti3Al/TiAl + TiC ceramic layer. In this study, TiC particle-dispersed Ti3Al/TiAl matrix ceramic layer on the Ti-6Al-4V alloy by laser cladding has been researched by means of X-ray diffraction, scanning electron microscope, electron probe micro-analyzer, energy dispersive spectrometer. The main difference from the earlier reports is that Ti3Al/TiAl has been chosen as the matrix of the composite coating. The wear resistance of the Al + 30 wt.% TiC and the Al + 40 wt.% TiC cladding layer was approximately 2 times greater than that of the Ti-6Al-4V substrate due to the reinforcement of the Ti3Al/TiAl + TiC hard phases. However, when the TiC mass percent was above 40 wt.%, the thermal stress value was greater than the materials yield strength limit in the ceramic layer, the microcrack was present and its wear resistance decreased.

  4. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238PuO2-powered pacemaker could be transformed into a terrorism weapon

  5. Non-destructive Residual Stress Analysis Around The Weld-Joint of Fuel Cladding Materials of ZrNbMoGe Alloys

    Directory of Open Access Journals (Sweden)

    Parikin

    2003-08-01

    Full Text Available The residual stress measurements around weld-joint of ZrNbMoGe alloy have been carried out by using X-ray diffraction technique in PTBIN-BATAN. The research was performed to investigate the structure of a cladding material with high temperature corrosion resistance and good weldability. The equivalent composition of the specimens (in %wt. was 97.5%Zr1%Nb1%Mo½%Ge. Welding was carried out by using TIG (tungsten inert gas technique that completed butt-joint with a current 20 amperes. Three region tests were taken in specimen while diffraction scanning, While diffraction scanning, tests were performed on three regions, i.e., the weldcore, the heat-affected zone (HAZ and the base metal. The reference region was determined at the base metal to be compared with other regions of the specimen, in obtaining refinement structure parameters. Base metal, HAZ and weldcore were diffracted by X-ray, and lattice strain changes were calculated by using Rietveld analysis program. The results show that while the quantity of minor phases tend to increase in the direction from the base metal to the HAZ and to the weldcore, the quantity of the ZrGe phase in the HAZ is less than the quantity of the ZrMo2 phase due to tGe element evaporation. The residual stress behavior in the material shows that minor phases, i.e., Zr3Ge and ZrMo2, are more dominant than the Zr matrix. The Zr3Ge and ZrMo2 experienced sharp straining, while the Zr phase was weak-lined from HAZ to weldcore. The hydrostatic residual stress ( in around weld-joint of ZrNbMoGe alloy is compressive stress which has minimum value at about -2.73 GPa in weldcore region

  6. Non-destructive Residual Stress Analysis Around The Weld-Joint of Fuel Cladding Materials of ZrNbMoGe Alloys

    International Nuclear Information System (INIS)

    The residual stress measurements around weld-joint of ZrNbMoGe alloy have been carried out by using X-ray diffraction technique in PTBIN-BATAN. The research was performed to investigate the structure of a cladding material with high temperature corrosion resistance and good weldability. The equivalent composition of the specimens (in %wt.) was 97.5%Zr1%Nb1%Mo½%Ge. Welding was carried out by using TIG (tungsten inert gas) technique that completed butt-joint with a current 20 amperes. Three region tests were taken in specimen while diffraction scanning, While diffraction scanning, tests were performed on three regions, i.e., the weld core, the heat-affected zone (HAZ) and the base metal. The reference region was determined at the base metal to be compared with other regions of the specimen, in obtaining refinement structure parameters. Base metal, HAZ and weld core were diffracted by X-ray, and lattice strain changes were calculated by using Rietveld analysis program. The results show that while the quantity of minor phases tend to increase in the direction from the base metal to the HAZ and to the weld core, the quantity of the ZrGe phase in the HAZ is less than the quantity of the ZrMo2 phase due to Ge element evaporation. The residual stress behavior in material shows that minor phases i.e., Zr3Ge and ZrMo2 are more dominant than the Zr matrix. The Zr3Ge and ZrMo2 experienced sharp straining, while Zr phase was weak-lined from HAZ to weld core. The hydrostatic residual stress (σ) in around weld-joint of ZrNbMoGe alloy is compressive stress which has minimum value at about -2.73 GPa in weld core region. (author)

  7. Characterization of hydrogenation behavior on Mo-modified Zr-Nb alloys as nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    The effects of Mo in Zr-Nb alloys are investigated in terms of their mechanical properties associated with microstructure, as well as their behavior under hydrogen environment. Zr-Nb-Mo alloys were fabricated by arc melting and subsequently cold rolling and annealing below the eutectoid temperature. Hydrogen was absorbed in a furnace under argon and hydrogen gas flow environment at high temperature. X-Ray diffraction, electron backscatter diffraction, and tensile test were jointly utilized to carry out detailed microstructural characterization and mechanical properties. Results showed that fcc-δ-ZrH1.66 was formed in all hydrogen-absorbed alloys, and the amount of hydride enhanced with increasing of hydrogen content. In addition, it was clear that δ-ZrH1.66 was precipitated both in grain boundary and interior, and preferential precipitation was observed on the habit planes of (0001) and {101-bar7}. Moreover, the strengthening effect by Mo addition was observed. The ductility loss by hydrogen absorption was found from fracture surface observation. Large area cleavage facets were found in Mo-free specimen, and less cleavage facets was observed in Mo-containing specimen, showing an appropriate addition of Mo can increase the tolerance to hydrogen embrittlement. (author)

  8. Optimization of the basic parameters of cathodic deposition of Ce-conversion coatings on D16 am clad alloy

    International Nuclear Information System (INIS)

    Full text: The present research work is investigation on the probabilities for application of a new cerium compound, for cathodic electrodeposition of cerium based conversion coatings (CeCC) for protection of D16 AM alloy against corrosion. For the purpose of the present study, diammonium pentanitrocerate (NH4)2Ce(NO3)5 was used, where the cerium is represented in the anionic moiety, instead of the electrolytes used up to nowadays. The barrier ability and durability against corrosion of all coatings were evaluated by electrochemical methods - Linear Sweep Voltammetry (LSV) and Electrochemical Impedance Spectroscopy (EIS). Additionally, selected specimens underwent morphological characterization by means of Optical Metallographic Microscopy (OMM) and Scanning Electronic Microscopy (SEM) combined with Energy Dispersive X-ray spectroscopy (EDX). As a result, various parameters and conditions of deposition, such as the preliminary treatment, concentration of the basic substance and additives, density of the applied electric current and duration of deposition were elucidated. key words: corrosion protection, aluminium alloy CeCC, EIS, LSV, SEM, EDS

  9. Past research and fabrication conducted at SCK-CEN on ferritic ODS alloys used as cladding for FBR's fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    De Bremaecker, Anne, E-mail: adbremae@sckcen.be [Studiecentrum voor Kernenergie-Centre d' Etude de l' Energie Nucleaire (SCK-CEN), NMS, Mol (Belgium)

    2012-09-15

    and final reduction rates, temperature, duration, atmosphere and furnace). Specific non-destructive tests (ultrasonic and eddy currents) were also developed. In-pile creep in argon and in liquid sodium was deeply studied on pressurized segments irradiated up to 75 dpa{sub NRT}. Finally two fuel assemblies cladded with such ODS alloys were irradiated in Phenix to the max dose of 90 dpa. Creep deformation and swelling were limited but the irradiation-induced embrittlement became acute. The programme was stopped shortly after the Chernobyl disaster, before the embrittlement problem was solved.

  10. The plastic anisotropy of an Al-Li-Cu-Zr alloy extrusion in unidirectional deformation

    Science.gov (United States)

    Lyttle, M. T.; Wert, J. A.

    1996-11-01

    The plastic anisotropy resulting from the initial deformation microstructure and various aging treatments applied to several regions of an AA2090 near-net-shape extrusion has been investigated. Yield behavior was measured by uniaxial compression in multiple orientations of each region. Two models of the plastic anisotropy were generated: the Taylor/Bishop-Hill model, based on crystallographic texture, and the plastic inclusion model, developed by Hosford and Zeisloft,[5] which incorporates anisotropic-precipitate effects. In overaged conditions, the Taylor/Bishop-Hill model adequately describes the observed plastic anisotropy. As the strengthening increment due to second-phase particles increases, there is a concurrent increase in the magnitude of the precipitate contribution to anisotropy. This anisotropy can not be accurately predicted solely by crystallographic texture. By incorporation of terms describing the precipitate anisotropy, the plastic inclusion model correctly predicts the yield strength variation in all regions tested. Examination of the fundamental interaction between matrix and precipitation strengthening reveals that there is a stronger basis for taking the critical resolved shear stress (CRSS) of the precipitates as a constant, rather than their effective yield strength. This consideration provides a more consistent and accurate form of the plastic inclusion model.

  11. Characterization of fracture behavior of zirconium alloys for fuel rod cladding of nuclear power plant in the post-quench stage of a LOCA (Loss of Coolant Accident)

    International Nuclear Information System (INIS)

    In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100 C and 1200 C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle a(O) layer, and a layer which can have residual ductility - the inner ex-β layer. Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples. A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and a(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test. (author)

  12. Reactor physics assessment of alternate cladding materials

    International Nuclear Information System (INIS)

    A preliminary reactor physics assessment has been performed for candidate alternate cladding materials to replace zirconium alloys in enhanced accident tolerant fuel (ATF) concepts for light water reactors. Proposed ATF concepts seek to reduce severe accident risks by increasing the coping time available to operators for accident response and reducing the extent and rate of heat and hydrogen production from steam oxidation. Candidate materials in this neutronics-focused study included austenitic stainless steel 310SS, alumina-forming ferritic alloys (FeCrAl), and silicon carbide (SiC). Historic 304SS cladding and Zircaloy were considered as reference points. Initial results indicate that the metallic options require increased uranium enrichments and/or decreased cladding thicknesses to match the operating cycle lengths achieved with Zircaloy; FeCrAl offered the smallest reactivity penalty, whereas 310SS showed large negative impacts. Ceramic SiC cladding performed well if cladding thicknesses remained similar to those for Zircaloy, but large clad thickness increases led to negative impacts. Fuel pellet relative radial power distributions were similar for all clad materials analyzed. Finally, an economic assessment found that 310SS or FeCrAl could increase fuel pellet costs by 15–36%, while SiC fuel pellet costs were very similar to Zircaloy. (author)

  13. 铜-钢-铜三层复合板室温轧制成形工艺及结合机制的研究%The Study of the Compound Technology and Bond Mechanism Copper Alloy/Q195/Copper Alloy Three-ply Cladding Sheet

    Institute of Scientific and Technical Information of China (English)

    于宝义; 安振之; 齐克敏

    2001-01-01

    研究了铜合金-Q195-铜合金三层复合板室温轧制成形工艺,借助金相显微镜、扫描电镜、电子探针分析了复合板的结合机制。%The study was carried out on the compound technology of copper alloy/Q195/copper alloy cladding sheet,with room temperature rolling deformation.The bond mechanism of compound sheet was analysed by using optical,scanning electron microscope,electron probe.

  14. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

    2008-07-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  15. CPR1000核电蒸汽发生器管板镍基合金堆焊工艺改进%Improved Nickel -based Alloy Cladding Technology of Tube Sheet for CPR1000 Steam Generator

    Institute of Scientific and Technical Information of China (English)

    刘鸣宇; 吴绍炳; 吴义党

    2011-01-01

    The cladding with nickel - based alloys on steam generator tube sheet is the basic and key operation in the steam generator manufacturing processes. The cladding quality can affect the schedule and quality of follow - up operations directly. China Guangdong Nuclear CPR1000 project is the second generation nuclear power plant with the intellectual property of CNPEC. This article describes two mature technology concerning the steam generator tube sheet cladding of the CPR1000 nuclear power project in China. Based on above analysis, the improved technology is proposed . With the technology, the central region of tube sheet can be clad by automatic welding.%蒸汽发生器管板镍基堆焊是蒸汽发生器制造的基本工序,也是关键工序,堆焊质量直接影响到设备制造进度和后续工艺质量.CPR1000项目是中广核自主知识产权的二代加核电站,本文阐述了国内CPR1000核电项目蒸汽发生器管板堆焊较为成熟的两种工艺,在此基础上提出了改进管板堆焊的工艺方案,解决了CPR1000核电项目蒸汽发生器管板中心区域一定直径范围内无法实现自动堆焊的难题.

  16. Microstructure and Wear Resistance of Laser Cladding TiC Coat on Titanium Alloy%钛合金表面激光熔覆TiC涂层显微结构和耐磨性

    Institute of Scientific and Technical Information of China (English)

    王慧萍; 李军; 李芳; 李曼萍; 奚文龙

    2012-01-01

    采用HL-5000型横流CO2激光加工机在TC4钛合金表面激光熔覆TiC+ Ti和TiC+Ti+ F102复合涂层.通过SEM、EDAX、XRD、HXD-1000TMC型显微硬度计,HT-600型高温摩擦磨损试验机,分析了熔覆层的显微组织、成分、物相,测试了激光熔覆层的显微硬度和滑动摩擦磨损性能.结果表明,激光熔覆制备的TiC复合涂层与基体呈冶金结合,在TiC+ Ti激光熔覆层中,熔覆层的组织是在Ti基体上分布着TiC树枝晶;在TiC +Ti+ F102激光熔覆层中,TiC颗粒发生了部分溶解,熔覆层的组织是在Ti基和γ-Ni基的基体上分布着细小的TiC颗粒和TiC树枝晶.TiC+ Ti激光熔覆层的硬度约为700 HV0.1,TiC+Ti+ F102激光熔覆层的硬度约为800 HV0.1,两种复合涂层耐磨性均比TC4钛合金显著提高.%The laser cladding TiC + Ti and TiC + Ti + F102 composite coaling on the surfact of TC4 alloy was obtained with 5.0 Kw continuous wave CO2 laser. The microstructure,composition and phase of the coating were investigated by means of SEM,EDAX,XRD,HXD-1000TMC Microhardness Tester, HT-600 wear machine Moreover, the microhardness and friction wear properties was measured. The results indicate that the laser cladding TiC composite coating is well bonded with the matrix alloy. The microstructures of TiC dendrites in Ti matrix in the clad layer of TiC + Ti laser clad coating. For TiC + Ti + F102 laser clad coating, parts of TiC particles are dissolved to form a microstructures of TiC particles and fine TiC dendrites in the matrix of Ti and y-Ni in the clad layer. The microhardness of TiC + Ti coating is 700 HV0.1. The microhardness of TiC + Ti + F102 coating is 800 HV0.1 , and the coating greatly enhances the wear resistantce of TC4 titanium alloy.

  17. Characteristics of Ni-based coating layer formed by laser and plasma cladding processes

    Energy Technology Data Exchange (ETDEWEB)

    Xu Guojian [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: xuguojian1959@hotmail.com; Kutsuna, Muneharu [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: kutsuna@numse.nagoya-u.ac.jp; Liu Zhongjie [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: xyliuzhj8@hotmail.com; Zhang Hong [Changchun University of Science and Technology, 7 Weixing Road, Changchun, Jilin Province 130022 (China)]. E-mail: Zhanghongcust@hotmail.com

    2006-02-15

    The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO{sub 2} laser and plasma cladding processes. The smooth clad bead was obtained by CO{sub 2} laser cladding process. The phases of clad layer were investigated by an optical microscope, scanning electron microscopy (SEM), X-ray diffractometer (XRD), electron probe microanalysis (EPMA) and energy-dispersive spectrometer (EDS). The microstructures of clad layers belonged to a hypereutectic structure. Primary phases consist of boride CrB and carbide Cr{sub 7}C{sub 3}. The eutectic structure consists of Ni + CrB or Ni + Cr{sub 7}C{sub 3}. Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO{sub 2} laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality.

  18. Fatigue crack propagation of new aluminum lithium alloy bonded with titanium alloy strap

    Institute of Scientific and Technical Information of China (English)

    Sun Zhenqi; Huang Minghui

    2013-01-01

    A new type of aluminum lithium alloy (A1-Li alloy) Al-Li-S-4 was investigated by test in this paper.Alloy plate of 400 mm × 140 mm × 6 mm with single edge notch was made into samples bonded with Ti-6Al-4V alloy (Ti alloy) strap by FM 94 film adhesive after the surface was treated.Fatigue crack growth of samples was investigated under cyclic loading with stress ratio (R) of 0.1 and load amplitude constant.The results show that Al-Li alloy plate bonded with Ti alloy strap could retard fatigue crack propagation.Retardation effect is related with width and thickness of strap.Flaws have an observable effect on crack propagation direction.

  19. Application of Coating Technology for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured

  20. Microstructure Evolution and Wear Behavior of the Laser Cladded CoFeNi2V0.5Nb0.75 and CoFeNi2V0.5Nb High-Entropy Alloy Coatings

    Science.gov (United States)

    Jiang, Li; Wu, Wei; Cao, Zhiqiang; Deng, Dewei; Li, Tingju

    2016-04-01

    The high-entropy alloy (HEA) coatings have received considerable attentions owing to their unique structures and properties caused by the quick solidification. In this work, the CoFeNi2V0.5Nb0.75 and CoFeNi2V0.5Nb HEAs which show fully eutectic and hypereutectic microstructures in their casting samples were laser cladded on 304 stainless steel substrate with laser power of 1400, 1600, and 1800 W. Results show that the HEA coatings are composed of the FCC solid solution phase and the Fe2Nb-type Laves phase. The cladding zones of the CoFeNi2V0.5Nb0.75 and CoFeNi2V0.5Nb coatings show cellular dendritic crystals, while the bonding zones show directional columnar crystals. Compared to the 304 stainless steel substrate, the HEA coatings show better wear resistance because of the combination of the hard Fe2Nb-type Laves phase and the ductile FCC solid solution matrix. Moreover, the HEA coatings with power of 1600 W show the best wear resistance attributing to the maximum volume fraction of the hard Fe2Nb-type Laves phase.

  1. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    International Nuclear Information System (INIS)

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed

  2. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B. Jr.

    1999-10-21

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed.

  3. Increasing corrosion resistance of carbon steels by surface laser cladding

    Science.gov (United States)

    Polsky, V. I.; Yakushin, V. L.; Dzhumaev, P. S.; Petrovsky, V. N.; Safonov, D. V.

    2016-04-01

    This paper presents results of investigation of the microstructure, elemental composition and corrosion resistance of the samples of low-alloy steel widely used in the engineering, after the application of laser cladding. The level of corrosion damage and the corrosion mechanism of cladded steel samples were established. The corrosion rate and installed discharge observed at the total destruction of cladding were obtained. The regularities of structure formation in the application of different powder compositions were obtained. The optimal powder composition that prevents corrosion of samples of low-carbon low-alloy steel was established.

  4. Reaction Mechanisms in the Li3AlH6/LiBH4 and Al/LiBH4 Systems for Reversible Hydrogen Storage. Part 2: Solid-State NMR Studies

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Joon; Lu, Jun; Sohn, Hong Yong; Fang, Zhigang Zak; Kim, Chul; Bowman, Robert C.; Hwang, Son-Jong

    2011-09-01

    In Part 1, the promising hydrogen storage properties of the combined systems of Li3AlH6/LiBH4 and Al/LiBH4, exhibiting the favorable formation of AlB2 during dehydrogenation, were presented based on TGA and XRD analyses. The present Part 2 describes the characterization of the intermediate and final products of the dehydrogenation and rehydrogenation of the above systems by multinuclear solid state NMR characterization. This work has also verified that the presence of Al resulted in the re-formation of LiBH4 occurring at a much lower temperature and H2 pressure, under which conditions the dehydrogenation product from LiBH4 alone does not show any degree of rehydrogenation. NMR studies mainly identified various reaction intermediates for LiBH4 dehydrogenation/rehydrogenation reactions. Unlike the XRD studies, the AlB2 formation in particular could not be unambiguously confirmed by NMR. 27Al NMR showed that aluminum was mainly involved in various Li-Al alloy formations. The catalytic role of Al in the LiBH4 hydrogen storage reactivity could be achieved by a reversible cycle of Al + LiH ↔ LiAl + 1/2H2 reaction.

  5. Clad metals by roll bonding for SOFC interconnects

    Science.gov (United States)

    Chen, L.; Jha, B.; Yang, Zhenguo; Xia, Guang-Guang; Stevenson, Jeffry W.; Singh, Prabhakar

    2006-08-01

    High-temperature oxidation-resistant alloys are currently considered as a candidate material for construction of interconnects in intermediate-temperature solid oxide fuel cells. Among these alloys, however, different groups of alloys demonstrate different advantages and disadvantages, and few, if any, can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, cladding has been proposed as one approach in fabricating metallic layered interconnect structures. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated. This paper provides a brief overview of the cladding approach and discusses the viability of this technology to fabricate the metallic layered-structure interconnects.

  6. Fracture Toughness Of Zircaloy Claddings

    International Nuclear Information System (INIS)

    Zirconium-based alloys (Zircaloy) have been used as cladding material in Light Water Reactors for many years. During fabrication, or in in-reactor service, crack-type defects can be formed, posing questions regarding mechanical integrity. As claddings change their mechanical properties (mainly toughness) during service as a result of irradiation-induced degradation, oxidation and hydride formation, it is essential for integrity considerations to provide parameters for the assessment of the influence of flaws on rupture behaviour. Usually, fracture-mechanics parameters are employed such as the fracture toughness, KIC, or, for high plastic strains, the J-integral, JIC. The applicability of these parameters is, however, limited by the dimensions of the samples (e.g. thickness). In claddings with a wall thickness of below 1 mm, determination of toughness necessitates an extension of the J-integral concept. A method based on the traditional J-approach, but applicable to thin-walled structures, is presented in this paper. (author)

  7. Inpile (in PWR) testing of cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs.

  8. A defect density-based constitutive crystal plasticity framework for modeling the plastic deformation of Fe-Cr-Al cladding alloys subsequent to irradiation

    International Nuclear Information System (INIS)

    It is essential to understand the deformation behavior of these Fe-Cr-Al alloys, in order to be able to develop models for predicting their mechanical response under varied loading conditions. Interaction of dislocations with the radiation-induced defects governs the crystallographic deformation mechanisms. A crystal plasticity framework is employed to model these mechanisms in Fe-Cr-Al alloys. This work builds on a previously developed defect density-based crystal plasticity model for bcc metals and alloys, with necessary modifications made to account for the defect substructure observed in Fe-Cr-Al alloys. The model is implemented in a Visco-Plastic Self Consistent (VPSC) framework, to predict the mechanical behavior under quasi-static loading.

  9. A defect density-based constitutive crystal plasticity framework for modeling the plastic deformation of Fe-Cr-Al cladding alloys subsequent to irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Patra, Anirban [Los Alamos National Laboratory; Wen, Wei [Los Alamos National Laboratory; Martinez Saez, Enrique [Los Alamos National Laboratory; Tome, Carlos [Los Alamos National Laboratory

    2016-02-05

    It is essential to understand the deformation behavior of these Fe-Cr-Al alloys, in order to be able to develop models for predicting their mechanical response under varied loading conditions. Interaction of dislocations with the radiation-induced defects governs the crystallographic deformation mechanisms. A crystal plasticity framework is employed to model these mechanisms in Fe-Cr-Al alloys. This work builds on a previously developed defect density-based crystal plasticity model for bcc metals and alloys, with necessary modifications made to account for the defect substructure observed in Fe-Cr-Al alloys. The model is implemented in a Visco-Plastic Self Consistent (VPSC) framework, to predict the mechanical behavior under quasi-static loading.

  10. Investigation on Wear Resistance of Fe-based Alloy Coating Prepared by Argon Arc Cladding on Q235%Q235钢氩弧熔覆铁基合金涂层的耐磨性研究

    Institute of Scientific and Technical Information of China (English)

    郭国林; 张娜; 王俊杰; 李刚

    2012-01-01

    和用氩弧熔覆技术,选择合适的工艺参数,在Q235钢材表面熔覆了铁基合金耐磨涂层.通过金相显微镜和SEM分析了熔覆涂层的显微组织,并测试了涂层的显微硬度和耐磨性.结果表明,在Q235钢表面制备了以马氏体组织和γ-(Fe-Cr-Ni-C)合金固溶体为基体,以(Cr,Fe)7C3、Fe3C、Fe2B等化合物为增强相的合金涂层;涂层的显微硬度可达600 HV;涂层的耐磨性较基体提高近8倍.在低碳钢表面熔覆一层耐磨材料,既保留了低碳钢较高的塑、韧性,又提高了表面层的硬度和耐磨性.%By plasma cladding technology,a wear resistant coating with Fe-based alloy was prepared on the surface of Q235 steel. The microstructure of the bonding coating was investigated by optical microscope and scanning electron microscope. The microhardness and wear resistance performance of the coating was tested. The results show that the alloy coating on the surface of Q235 is conposed of the matrix of martensitic structure and y- (Fe-Cr-Ni-C) alloy solid solution ,and the reinforcing phases of (Cr, Fe)7Cγ/Fe3C/Fe2B and other compounds. The microhardness of the coating can reach 600 HV and the wear resistance is about 8 times higher than that of Q235 steel substrate. When deposited a wear-resistant layei on the surface of mild steel, the high plasticity and ductility of mild steel can be preserved, and the hardness and wear resistance of the cladding layer can also be improved greatly.

  11. POROSITY DEVELOPMENT DURING HEAT TREATMENT OF ALUMINUM-LITHIUM ALLOYS

    OpenAIRE

    Papazian, J.; J. Wagner; Rooney, W.

    1987-01-01

    The development of a sub-surface layer of porosity during heat treatment has been studied in a variety of Al-Li alloys. Pure binary Al-Li alloys and three commercial materials were heat treated in air, vacuum and hydrogen for various lengths of time. Subsequent metallographic sectioning and polishing revealed the presence of a band of pores in the near-surface region extending approximately 300 µm into the sample after a 16 h heat treatment. This band of porosity is thought to arise from a Ki...

  12. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  13. Laser cladding with wide-band scanning rotative polygon mirror

    International Nuclear Information System (INIS)

    This paper discusses the scanning rotative polygon mirror providing a uniform linear heat source with both amplitude and frequency continuous adjustment that has been developed to produce singlepass widths about 14mm and 13mm, fourpass widths about 43mm and 35mm respectively for NiCrSiB and FeCrSiB alloy cladded on A3 substrate. Bead side angles were 175 degrees and 167 degrees respectively above alloys. A very large smooth area with average roughness Ra = 0.64μm was made by NiCrSiB alloy laser cladded

  14. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  15. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  16. Corrosion protection of clad 2024 aluminum alloy anodized in tartaric-sulfuric acid bath and protected with hybrid sol–gel coating

    International Nuclear Information System (INIS)

    Clad AA2024 T3 specimens were anodized in a tartaric-sulfuric acid bath (TSA) and subsequently protected either by classical Cr-free water sealing treatment or by application of a hybrid sol–gel coating. The sol–gel coating was prepared using a solution with high water content (58 %v/v) and obtained by the hydrolysis and condensation of tetraethoxysilane (TEOS) and 3-glycidoxypropyltrimethoxysilane (GPTMS). The morphology of the sealed anodic films and their thicknesses were evaluated by scanning electron microscopy (SEM), field emission scanning electron microscopy (FE-SEM) and glow discharge optical emission spectrometry (GDOES). The corrosion resistance of the samples was evaluated by electrochemical impedance spectroscopy (EIS) and salt-spray test. The results showed that the treatment with the hybrid sol–gel increased the resistive properties of the pores compared to the classical water sealing, delaying the access of aggressive species to the barrier layer

  17. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria

  18. Mechanical alloying of aluminium-lithium-magnesium alloy powders

    International Nuclear Information System (INIS)

    The production of high-purity aluminium-lithium-magnesium alloy powders, by mechanical alloying through grinding in a vibratory mill under high vacuum at room temperature, is described in details. The source materials for the grinding mixture were: aluminium-lithium alloy powder obtained by thermal vacuum-dehydrogenization of AlLiH4 hydride; magnesium metal powder; and chemically deoxidized aluminium metal powder. The implications which arose from the high reactivity of the component elements are discussed, and the measures taken to overcome them are described. The procedures used for the chemical analysis and powder characterization are given. (orig.)

  19. Fuel cladding tubes and fuel elements

    International Nuclear Information System (INIS)

    Purpose: To enable non-destructive measurement for the thickness of zirconium barriers. Constitution: Regions capable of non-destructive inspection are provided at the boundary between a fuel cladding tube made of zirconium alloy and the zirconium barrier lined to the inner circumference surface of the tube. As the regions being capable of distinguishing by ultrasonic wave reflection, solid materials, for example, non-metal materials different from that for the tube and the barrier are placed or gaps are provided at the boundary between the zirconium alloy cladding tube and the zirconium barrier. Since ultrasonic waves are reflected at each of the boundaries by the presence of these regions, thickness of the zirconium barrier can be measured in a non-destructive manner from either the inner or the outer surface of the tube. (Yoshino, Y.)

  20. Experimental research of irradiated nuclear fuel cladding failure processes: OECD Studsvik Cladding Integrity Project II

    International Nuclear Information System (INIS)

    The following 4 partial tasks were addressed: V001: Experimental results and knowledge of the effect of the material properties of the cladding and pellet on the phenomena of mechanical fuel-cladding interaction under the effect of radiation, at different temperatures and RAMP power load; V002: Knowledge based on the analysis of experimental data concerning the effect of iodine on the development of cracks on the fuel pin cladding tubes; V003: Processing the results of experiments to determine the primary cause of delayed hydride cracking (DHC) initiation in modern cladding alloys with low hydrogen concentrations; and V004: Analysis of the result of research into the effect of hydrides and hydrogen in the solid solution on the extension of nuclear fuel pin cladding. The results corroborated the prediction capabilities of the FEMAXI-6 code. The calculations were performed both for the reactor ramp tests and for the relaxation tests of the cladding materials, where MKP SW was the dominant tool. MKP was used for calculations within the bilateral relations with Studsvik Nuclear in the preparation of a new mechanical test for investigation of DHC, and basic MKP analyses were performed for the off-reactor test with an expansion mandrel. The theoretical generalization of the unique experimental data is documented through analysis and description of the final validation phase within the Quantum Technologies MKP model. (P.A.)

  1. Special techniques for tensile tests of irradiated zirconium claddings

    International Nuclear Information System (INIS)

    Irradiated zirconium alloy claddings possessing property anisotropy should be tested in transverse and longitudinal directions. Such mechanical tests can be performed in conditions of large variety of geometric peculiarities of specimens, supports or grips. The objective of the work is the development of the unified complex of updated special techniques that allow investigation of mechanical pre- and post-irradiation properties of VVER claddings including radiation effect of property anisotropy changes in the same way. (author)

  2. Investigation of laser cladding high temperature anti-wear composite coatings on Ti6Al4V alloy with the addition of self-lubricant CaF{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Xiang, Zhan-Feng [School of Mechanical and Electric Engineering, Soochow University, 178 East Ganjiang Road, Suzhou 215006 (China); Liu, Xiu-Bo, E-mail: liuxiubo@suda.edu.cn [School of Mechanical and Electric Engineering, Soochow University, 178 East Ganjiang Road, Suzhou 215006 (China); Ren, Jia; Luo, Jian; Shi, Shi-Hong; Chen, Yao [School of Mechanical and Electric Engineering, Soochow University, 178 East Ganjiang Road, Suzhou 215006 (China); Shi, Gao-Lian; Wu, Shao-Hua [Suzhou Institute of Industrial Technology, Suzhou 215104 (China)

    2014-09-15

    Highlights: • A novel high temperature self-lubricating wear-resistant coating was fabricated. • TiC carbides and self-lubricant CaF{sub 2} were “in situ” synthesized in the coating. • The coating with the addition of CaF{sub 2} possessed superior properties than without. - Abstract: To improve the high-temperature tribological properties of Ti–6Al–4V alloy, γ-NiCrAlTi/TiC and γ-NiCrAlTi/TiC/CaF{sub 2} coatings were fabricated on Ti–6Al–4V alloy by laser cladding. The phase compositions and microstructure of the coatings were investigated using X-ray diffraction (XRD) and scanning electron microscope (SEM) equipped with energy dispersive spectroscopy (EDS). The tribological behaviors were evaluated using a ball-on-disk tribometer from ambient temperature to 600 °C under dry sliding wear conditions and the corresponding wear mechanisms were discussed. The results indicated that the γ-NiCrAlTi/TiC/CaF{sub 2} coating consisted of α-Ti, the “in situ” synthesized TiC block particles and dendrite, γ-NiCrAlTi solid solution and spherical CaF{sub 2} particles. The wear rates of γ-NiCrAlTi/TiC/CaF{sub 2} coating were decreased greatly owing to the combined effects of the reinforced carbides and continuous lubricating films. Furthermore, the friction coefficients of γ-NiCrAlTi/TiC/CaF{sub 2} coating presented minimum value of 0.21 at 600 °C, which was reduced by 43% and 50% compared to the substrate and γ-NiCrAlTi/TiC coating respectively. It was considered that the γ-NiCrAlTi/TiC/CaF{sub 2} coating exhibited excellent friction-reducing and anti-wear properties at high temperature.

  3. Investigation of laser cladding high temperature anti-wear composite coatings on Ti6Al4V alloy with the addition of self-lubricant CaF2

    International Nuclear Information System (INIS)

    Highlights: • A novel high temperature self-lubricating wear-resistant coating was fabricated. • TiC carbides and self-lubricant CaF2 were “in situ” synthesized in the coating. • The coating with the addition of CaF2 possessed superior properties than without. - Abstract: To improve the high-temperature tribological properties of Ti–6Al–4V alloy, γ-NiCrAlTi/TiC and γ-NiCrAlTi/TiC/CaF2 coatings were fabricated on Ti–6Al–4V alloy by laser cladding. The phase compositions and microstructure of the coatings were investigated using X-ray diffraction (XRD) and scanning electron microscope (SEM) equipped with energy dispersive spectroscopy (EDS). The tribological behaviors were evaluated using a ball-on-disk tribometer from ambient temperature to 600 °C under dry sliding wear conditions and the corresponding wear mechanisms were discussed. The results indicated that the γ-NiCrAlTi/TiC/CaF2 coating consisted of α-Ti, the “in situ” synthesized TiC block particles and dendrite, γ-NiCrAlTi solid solution and spherical CaF2 particles. The wear rates of γ-NiCrAlTi/TiC/CaF2 coating were decreased greatly owing to the combined effects of the reinforced carbides and continuous lubricating films. Furthermore, the friction coefficients of γ-NiCrAlTi/TiC/CaF2 coating presented minimum value of 0.21 at 600 °C, which was reduced by 43% and 50% compared to the substrate and γ-NiCrAlTi/TiC coating respectively. It was considered that the γ-NiCrAlTi/TiC/CaF2 coating exhibited excellent friction-reducing and anti-wear properties at high temperature

  4. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes a series of the characterization results of candidate alloys, the manufacturing description of the advanced sample cladding tubes, and both the summary of out-of pile tests and the overview plan of in-pile test for them. Ten(10) kinds of the second candidate alloys, which had been selected at the first stage of the project, were comprehensively tested for their out-of pile performance. Six(6) kinds of the alloys were selected of the second ones as the final candidates through the screening tests. The out-of pile performance of the final candidates were superior to that of zircaloy-4. The advanced sample cladding tubes were made of the final candidates and tested for their out-of pile performances. The corrosion behaviors of the tubes were evaluated though the corrosion tests in water at 360 .deg. C, steam at 400 .deg. C and LiOH solution at 360 .deg. C. The mechanical properties such as creep, tensile and burst were also evaluated for each tube. The textures, microstructures, precipitates and hydrides of each tube were analyzed as well as the phase transformation was studied for each tube. In general, the test results showed that the performance of the advanced sample cladding tubes was improved over 30% in corrosion and 20% in mechanical property than that of zircaloy-4. The in-pile test of the tubes for the first phase was arranged from January 2003 to March 2007

  5. Laser cladding of cobalt and boron free hardfacing materials for nuclear applications

    International Nuclear Information System (INIS)

    Full text: Most common hardfacing alloys are of stellite family, which are cobalt base alloys and borides are also being developed for the same purpose. Both cobalt and boron are not preferred in nuclear industry as cobalt becomes radioactive after irradiation and boron is a neutron poison. Therefore, there is a need to develop cobalt and boron free hardfacing alloys for nuclear application. Attempt has been made to develop cobalt - boron free hardfacing alloys for laser cladding. Laser cladding of three nickel based hardfacing materials, one metallic system (Ni-15Cr-32Mo) and two composite systems (Ni-20Cr)-40Cr2C3 and (Ni-20Cr)-40WC) has been attempted. These hardfacing materials were cladded onto 0.15C steel sheet by blown powder laser cladding. Laser cladding of stellite was also done for comparative purpose. The process parameters were optimised to obtain defect free cladding. The cladded samples were characterized by visual, optical microscopy and microhardness measurements. Wear testing of these claddings was done by the pin on disc method against 600-grit size abrasive paper. Comparative study of wear properties of these claddings was done. Results of these investigations are reported in this paper

  6. Modeling alternative clad behavior for accident tolerant systems

    International Nuclear Information System (INIS)

    The US Department of Energy Fuel Cycle Research and Development program has a key goal of helping develop accident tolerant fuels (ATF) through investigating fuel and clad forms. In the current work thermochemical modeling and experiment are being used to assess fuel and clad alternatives. Cladding alternatives that have promise to improve fuel performance under accident conditions include the FeCrAl family of alloys and SiC-based composites. These are high strength and radiation resistant alloys and ceramics that have increased resistance to oxidation as compared to zirconium alloys. Accident modeling codes have indicated substantially increased time to failure and resulting effects. In the current work the thermochemical behavior of these materials are being assessed and the work reported here. (author)

  7. Study on modes of energy action in laser-induction hybrid cladding

    International Nuclear Information System (INIS)

    The shape and microstructure in laser-induction hybrid cladding were investigated, in which the cladding material was provided by means of three different methods including the powder feeding, cold pre-placed coating (CPPC) and thermal pre-placed coating (TPPC). Moreover, the modes of energy action in laser-induction hybrid cladding were also studied. The results indicate that the cladding material supplying method has an important influence on the shape and microstructure of coating. The influence is decided by the mode of energy action in laser-induction hybrid cladding. During the TPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating. During the CPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating and substrate surface, respectively. In powder feeding hybrid cladding, a part of laser is absorbed by the powder particles directly, while the other part of laser penetrating powder cloud radiates on the molten pool. Meanwhile, the induction heating is entirely performed on the substrate. In addition, the wetting property on the interface is improved and the metallurgical bond between the coating and substrate is much easier to form. Therefore, the powder feeding laser-induction hybrid cladding has the highest cladding efficiency and the best bond property among three hybrid cladding methods.

  8. Zircaloy-4 hydriding. Hydrogen distribution in PWR's rod cladding

    International Nuclear Information System (INIS)

    In pressurised water reactors, Zircaloy 4 is used as fuel cladding in contact with hot water. The precipitation of hydrides at room temperatures causes mechanical deterioration of the cladding. As the cladding is subjected to a radial temperature gradient, the hydrogen distribution is greatly affected. The image analysis method is used to determine the hydride distribution in the irradiated cladding. To calibrate this method, a device was specially built for the preparation of Zircaloy specimens with known hydrogen contents. The hydriding conditions and hydrogen content determination procedures were fixed. We have successfully realized specimens with various hydrogen contents. With these specimens, a relationship between the parameter Sv (surface density of hydrides) and the hydrogen content was established. This parameter Sv is independent from the Zircaloy 4 metallurgical state (i.e. stress relieved or recrystallized) and from the analysis section (longitudinal or transverse). Study of hydrogen content and hydride distribution in irradiated cladding by means of image analysis showed that the method is limited by its ability of separation between neighbouring hydrides at cladding's periphery where the hydrogen content can reach several thousands ppm. Nevertheless, this method gives us some information about hydride distribution inside the cladding. A model for thermal diffusion was developped to stimulate the migration of hydrogen in Zirconium alloys. This model was used to predict hydrogen distribution in the irradiated cladding. Comparison of model predictions with results of image analysis shows good agreement. (Author). refs., figs., tabs

  9. Mechanical interaction fuel/cladding

    International Nuclear Information System (INIS)

    There is a common agreement that rather large plastic cladding deformation may occur in fast breeder reactor conditions. In thermal irradiation experiments these deformations can be directly measured as cladding diameter increase. In case of fast flux, a distinction must be made between plastic strain and swelling due to pore formation. The separation of these two effects can be made by a combination of cladding diameter measurements and cladding density measurements. A simpler method to determine the mean plastic cladding expansion is to compare the increase of relative mean cladding diameter along the fuel element and the increase of relative cladding length. This comparison for the irradiation experiment in Rapsodie is shown

  10. Preparation, microstructural evolution and properties of Ni–Zr intermetallic/Zr–Si ceramic reinforced composite coatings on zirconium alloy by laser cladding

    International Nuclear Information System (INIS)

    NiZr2–ZrSi–Zr5(SixNi1−x)4-ZrC intermetallic/ceramic reinforced composite coatings were in situ synthesized by laser cladding the pre-placed Ni–Cr–B–Si powder on zirconium substrate. Microstructure and phase constituents were investigated by X-ray diffraction (XRD), optical microscope (OM), scanning electron microscope (SEM) and energy dispersive spectroscopy (EDS). Microhardness tester and block-on-ring wear tester were employed to measure the hardness distribution and wear resistance of the intermetallic/ceramic reinforced composite coating. Results indicated that the multiphase of reinforcements includes Ni–Zr intermetallic compounds (e.g., NiZr and NiZr2) and Zr–Si(C) ceramic phases (e.g., ZiSi, Zr5Si4 and ZrC). Ni–Si clusters transforming to Zr–Si–Ni clusters at high temperature facilitated the forming of Zr5(SixNi1−x)4 and during the growth of Zr5(SixNi1−x)4, the consumption of Zr atoms at the lateral interface of liquid/Zr5(SixNi1−x)4 resulted into developing Zr-poor zone near Zr5(SixNi1−x)4. The microhardness and wear resistance of the coating were significantly improved by various reinforced phases in comparison to zirconium substrate. - Highlights: • NiZr2–ZrSi–Zr5(SixNi1−x)4-ZrC compostie coating was in-situ synthesized. • Ni–Si clusters transforming resulted into developing Zr-poor zone near Zr5(SixNi1−x)4. • Reinforced phases significantly improve wear resistance of the coating

  11. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  12. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    International Nuclear Information System (INIS)

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding

  13. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  14. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  15. A coating to protect spent aluminium-clad research reactor fuel assemblies during extended wet storage

    International Nuclear Information System (INIS)

    Pitting corrosion of aluminium (Al) alloy clad research reactor (RR) fuel in wet storage facilities can be reduced to a large extent by maintaining water parameters within specified limits. However, factors like bimetallic contact, settled solids and synergistic effects of many storage basin water parameters provoke cladding corrosion. Increase in corrosion resistance of spent Al-clad RR fuels can be achieved through the use of conversion coatings. This paper presents: (a) details about the formation of cerium dioxide as a conversion coating on Al alloys used as RR fuel cladding; (b) the corrosion resistance of cerium dioxide coated Al alloy specimens exposed to NaCl solutions. Marked improvements in corrosion resistance of cerium dioxide coated Al specimens were observed. This paper also presents details of a Latin American Project to develop conversion coatings for long term safe wet storage of spent Al-clad RR spent fuel assemblies. (author)

  16. Choice of methods and determination of fracture toughness for anticorrosion cladding metal

    International Nuclear Information System (INIS)

    Technique for fracture toughness determination within wide temperature range is chosen. Experiment results on austenitic anticorrosion cladding metal cracking resistance are given in comparison with temperature dependence low envelope of 15Ch2MFA steel fracture toughness. From the data obtained it follows, that crack propagation direction along cladding metal does not affect KIJ fracture toughness value. It is shown, that fracture toughness values of anticorrosion layer material are higher, than those of low-alloy steel for cladding

  17. Preparation, microstructural evolution and properties of Ni–Zr intermetallic/Zr–Si ceramic reinforced composite coatings on zirconium alloy by laser cladding

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Kun; Li, Yajiang, E-mail: yajli@sdu.edu.cn; Wang, Juan; Ma, Qunshuang; Li, Jishuai; Li, Xinyue

    2015-10-25

    NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC intermetallic/ceramic reinforced composite coatings were in situ synthesized by laser cladding the pre-placed Ni–Cr–B–Si powder on zirconium substrate. Microstructure and phase constituents were investigated by X-ray diffraction (XRD), optical microscope (OM), scanning electron microscope (SEM) and energy dispersive spectroscopy (EDS). Microhardness tester and block-on-ring wear tester were employed to measure the hardness distribution and wear resistance of the intermetallic/ceramic reinforced composite coating. Results indicated that the multiphase of reinforcements includes Ni–Zr intermetallic compounds (e.g., NiZr and NiZr{sub 2}) and Zr–Si(C) ceramic phases (e.g., ZiSi, Zr{sub 5}Si{sub 4} and ZrC). Ni–Si clusters transforming to Zr–Si–Ni clusters at high temperature facilitated the forming of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} and during the growth of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}, the consumption of Zr atoms at the lateral interface of liquid/Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. The microhardness and wear resistance of the coating were significantly improved by various reinforced phases in comparison to zirconium substrate. - Highlights: • NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC compostie coating was in-situ synthesized. • Ni–Si clusters transforming resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. • Reinforced phases significantly improve wear resistance of the coating.

  18. Effects of high temperature treatment on microstructure and mechanical properties of laser-clad NiCrBSi/WC coatings on titanium alloy substrate

    Energy Technology Data Exchange (ETDEWEB)

    Li, Guang Jie; Li, Jun, E-mail: jacob_lijun@sina.com; Luo, Xing

    2014-12-15

    Laser-clad composite coatings on the Ti6Al4V substrate were heat-treated at 700, 800, and 900 °C for 1 h. The effects of post-heat treatment on the microstructure, microhardness, and fracture toughness of the coatings were investigated by scanning electron microscopy, X-ray diffractometry, energy dispersive spectroscopy, and optical microscopy. The wear resistance of the coatings was evaluated under dry reciprocating sliding friction at room temperature. The coatings mainly comprised some coarse gray blocky (W,Ti)C particles accompanied by the fine white WC particles, a large number of black TiC cellular/dendrites, and the matrix composed of NiTi and Ni{sub 3}Ti; some unknown rich Ni- and Ti-rich particles with sizes ranging from 10 nm to 50 nm were precipitated and uniformly distributed in the Ni{sub 3}Ti phase to form a thin granular layer after heat treatment at 700 °C. The granular layer spread from the edge toward the center of the Ni{sub 3}Ti phase with increasing temperature. A large number of fine equiaxed Cr{sub 23}C{sub 6} particles with 0.2–0.5 μm sizes were observed around the edges of the NiTi supersaturated solid solution when the temperature was further increased to 900 °C. The microhardness and fracture toughness of the coatings were improved with increased temperature due to the dispersion-strengthening effect of the precipitates. Dominant wear mechanisms for all the coatings included abrasive and delamination wear. The post-heat treatment not only reduced wear volume and friction coefficient, but also decreased cracking susceptibility during sliding friction. Comparatively speaking, the heat-treated coating at 900 °C presented the most excellent wear resistance. - Highlights: • TiC + WC reinforced intermetallic compound matrix composite coatings were produced. • The formation mechanism of the reinforcements was analyzed. • Two precipitates were generated at elevated temperature. • Cracking susceptibility and microhardness of the

  19. Effects of the inner mould material on the aluminium–316L stainless steel explosive clad pipe

    International Nuclear Information System (INIS)

    Highlights: ► Different mould materials were adopted to evaluate the effect of the constraint on the clad quality. ► The interface characteristics of clad pipe were analyzed for the different clad pipe. ► The clad pipes possess excellent bonding quality. - Abstract: The clad pipe played an important part in the pipeline system of the nuclear power industry. To prepare the clad pipe with even macrosize and excellent bonding quality, in this work, different mould materials were adopted to evaluate the effect of the constraint on the clad quality of the bimetal pipe prepared by explosive cladding. The experiment results indicated that, the dimension uniformity and bonding interface of clad pipe were poor by using low melting point alloy as mould material; the local bulge or the cracking of the clad pipe existed when the SiC powder was utilized. When the steel mould was adopted, the outer diameter of the clad pipe was uniform from head to tail. In addition, the metallurgical bonding was formed. Furthermore, the results of shear test, bending test and flattening test showed that the bonding quality was excellent. Therefore, the Al–316L SS clad pipe could endure the second plastic forming

  20. Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    McClelland, R.G.; O' Leary, P.M. (Siemens Nuclear Power Corp., Richland, WA (United States))

    1992-01-01

    Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an [approximately]0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4.

  1. Madelung energy for random metallic alloys in the coherent potential approximation

    DEFF Research Database (Denmark)

    Korzhavyi, P. A.; Ruban, Andrei; Abrikosov, I. A.;

    1995-01-01

    one to include charge-transfer effects in the framework of the CPA. We show how the models work in actual calculations for selected metallic alloy systems, Al-Li, Li-Mg, and Ni-Pt, which exhibit charge transfer. We find that the so-called screened impurity model (β=1), which is derived completely...... within the mean-field single-site approximation, leads to the best agreement with experimental lattice parameter and mixing energy data for Al-Li and Li-Mg alloys. However, for the Ni-Pt system exhibiting strong ordering tendency this model seems to overestimate the Madelung energy of the completely...

  2. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  3. Material Selection for Accident Tolerant Fuel Cladding

    Science.gov (United States)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as >100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥1473 K (1200 °C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases, and FeCrAl alloys. Recently reported low-mass losses for Mo in steam at 1073 K (800 °C) could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1473 K (1200 °C) in steam and significant TiO2, and therefore, Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1748 K (1475 °C), while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at >1673 K (1400 °C) are still being evaluated.

  4. Initial Cladding Condition

    International Nuclear Information System (INIS)

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  5. Initial Cladding Condition

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2000-08-22

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis

  6. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  7. Cladding of pressure vessel steel with corrosion resistant filler material

    International Nuclear Information System (INIS)

    Pressure vessels are often on the inside clad with corrosion resistant material. Of the various cladding processes surfacing by welding has proved to be most useful, especially for large thick-walled pressure vessels. Submerged arc welding with strip electrode is the most common method. Rather promising results have also been obtained by plasma hot wire welding. In general, Nb-alloyed austenitic stainless steel, over-alloyed with Cr and Ni, is used as filler material. Henceforth, also nickel alloys, e.g. Inconel 600, are used. The surfacing is made in one or several layers, following the requirements on the clad surface and the welding process used. The most dangerous welding defects in the surface are various types of cracks. The corrosion resistance of the cladding can show rather high local variations, depending on the composition of the filler material and various welding process factors. It is proved that the surface layer comparises areas with low chromium martensite. To ensure the corrosion resistance of the cladding, the generation of low-chromium martensite must be prevented by using suitable welding parameters, welding equipment and filler metal. It is also possible to eliminate the negative influence on the corrosion resistance from the low-chromium martensite, e.g. by welding in two layers. In the case of the high demands on quality a welding procedure test should always be made prior to production welding.(author)

  8. Effects of solution heat treatment on the microstructure and hardness of Mg-5Li-3Al-2Zn-2Cu alloy

    International Nuclear Information System (INIS)

    The microstructure and hardness of Mg-5Li-3Al-2Zn-2Cu alloy were investigated both in the as-cast condition and after solution heat treatment at 330-390 deg. C for 5 h. The as-cast alloy contains a microstructure consisting of α-Mg matrix, AlLi phase, AlCuMg phase and Al2Cu phase. After the solution heat treament, the AlLi phase was dissolved into the matrix, however, the AlCuMg and Al2Cu phases were not dissolved. With the increase of solution temperature, almost all the AlLi phase was dissolved, and the effects of solution strengthening of Al and Li atoms in the alloy increase, which results in the gradual increase of the Brinell hardness of the solution-treated alloy.

  9. Electrochemical profiling of multi-clad aluminium sheets used in automotive heat exchangers

    OpenAIRE

    Bordo, Kirill; Ambat, Rajan; Peguet, Lionel; Afseth, Andreas

    2014-01-01

    The objective of the present study is to understand the mechanisms of corrosion propagation across the multi-clad structure of Al alloys sheets as a function of local alloy composition and microstructure, with and without brazing treatment. Electro-chemical behaviour at different depths was profiled using a combination of glow dis-charge optical emission spectroscopy (GDOES) sputtering, localized potentiodynam-ic polarization and zero resistance ammetry (ZRA) measurements. Multi-clad struc-tu...

  10. Proposal of 99.99%-aluminum/7N01-Aluminum clad beam tube for high energy booster of Superconducting Super Collider

    International Nuclear Information System (INIS)

    Proposal of 99.99% pure aluminum/7N01 aluminum alloy clad beam tube for high energy booster in Superconducting Super Collider is described. This aluminum clad beam tube has many good performances, but a eddy current effect is large in superconducting magnet quench collapse. The quench test result for aluminum clad beam tube is basically no problem against magnet quench collapse. (author)

  11. Development of Co-Pilgering Process for Manufacturing Double Clad Tubes for Accident Tolerant Fuel

    International Nuclear Information System (INIS)

    Accident Tolerant Fuels (ATF) are those that, in comparison with the standard UO2 - Zr system, can tolerate loss of active cooling in the core for a considerably longer time period (depending on the accident scenario), while maintaining or improving the fuel performance during normal operations. ATF cladding development efforts focus on materials with more benign steam reaction. For this, advanced steels (e.g. FeCrAl), refractory metals (e.g. Mo), ceramic cladding (SiC), Innovative alloys with dopants, zirconium alloy with coating or sleeve are being developed. Single material like zirconium alloy as clad may not be compatible with both fuel and coolant at elevated temperatures in accident scenario. Double clad tube is one of the prime concepts which has to be explored to develop ATF cladding. Two different clad materials- one oxidant resistant (like FeCrAl) and the other, fuel compatible (like Zr-4) constitute together as outer and inner tube to form ATF cladding. Bonding two different tubes in controlled thickness ratios and with almost no gap in between is utmost difficult. Different types of processes are available for production of double clad tubes such as coating, co-extrusion, co- drawing, internal expansion/external compaction, explosive bonding, co-pilgering etc,. Nuclear Fuel Complex (NFC), India has successfully demonstrated manufacturing of double clad tube by co-pilgering process where in outer cladding is of modified 9Cr-1Mo Steel and inner liner is of zircaloy-4. Considering different deformation behaviour of above materials during pilgering, fabrication of double clad tube is very critical. Optimization of tube dimensions like outer diameter and wall thickness at pre and final stages during pilgering is very important to achieve the required overall tube dimension and bonding between the tubes. This paper gives the methodology of manufacture of Double Clad Tubes by pilgering and the bonding between the two materials achieved in this process

  12. Cladding and wrapper development for fast breeder reactor high performance

    International Nuclear Information System (INIS)

    In order to ensure economic performance, of both the existing reactors and the future EFR, much recent research has been carried out within the framework of the European R and D agreement to examine the properties of various wrapper and cladding alloys. This paper reviews the status of the European research and development programmes on these steels and highlights the most striking results. For the cladding alloys, results on dimensional stability and tensile properties for fuel pin cladding irradiated in PFR or Phenix will be given. As for wrappers the presently available results of those wrappers irradiated in Phenix and PFR show that both ferritic steels are very good candidates and that on the basis of our present knowledge most of the properties are satisfactory for wrapper applications

  13. Development of powder metallurgy Al alloys for high temperature aircraft structural applications, phase 2

    Science.gov (United States)

    Chellman, D. J.

    1982-01-01

    In this continuing study, the development of mechanically alloyed heat resistant aluminum alloys for aircraft were studied to develop higher strength targets and higher service temperatures. The use of higher alloy additions to MA Al-Fe-Co alloys, employment of prealloyed starting materials, and higher extrusion temperatures were investigated. While the MA Al-Fe-Co alloys exhibited good retention of strength and ductility properties at elevated temperatures and excellent stability of properties after 1000 hour exposure at elevated temperatures, a sensitivity of this system to low extrusion strain rates adversely affected the level of strength achieved. MA alloys in the Al-Li family showed excellent notched toughness and property stability after long time exposures at elevated temperatures. A loss of Li during processing and the higher extrusion temperature 482 K (900 F) resulted in low mechanical strengths. Subsequent hot and cold working of the MA Al-Li had only a mild influence on properties.

  14. Advances in aluminium alloy products for structural applications in transportation

    OpenAIRE

    Staley, J.; Lege, D.

    1993-01-01

    This paper describes the needs of the aviation and automotive markets for structural materials and presents examples of developments of aluminum alloy products to fill these needs. Designers of aircraft desire materials which will allow them to design lightweight, cost-effective structures which have the performance characteristics of durability and damage tolerance. Their needs are being met by new and emerging materials varying from Al-Li alloys for thick structure, high-strength plate and ...

  15. In-pile cladding tests at NRI Rez and PIE capabilities and experience

    International Nuclear Information System (INIS)

    In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

  16. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  17. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made

  18. Effects of the manufacturing parameter and chemical composition on properties of HANA-4 cladding tube

    International Nuclear Information System (INIS)

    KEPCO NF conducted some researches to improve workability of HANA-4 cladding tube. It was changed to TREX outer diameter for increase Q-factor in first pilgering process related to the workability of cladding tube. In general, a increasing Q-factor leads to improvement yield of tubing manufacture in zirconium alloys. And decreasing of amount of alloying element changed cladding properties. The secondary phase particle analysis, the corrosion behavior and the texture were examined for HANA-4 alloys with adjustments of chemical compositions and TREX outer diameter for the purpose of enhancement formability. The precipitate type, size, and distribution of HANA-4 alloy were not changed as the chemical composition and the manufacturing parameters. The corrosion weight gain was decreased with reducing alloying elements, which considered the beneficial effect of reduced tin

  19. Analysis and optimization of process parameters in Al-SiCp laser cladding

    Science.gov (United States)

    Riquelme, Ainhoa; Rodrigo, Pilar; Escalera-Rodríguez, María Dolores; Rams, Joaquín

    2016-03-01

    The laser cladding process parameters have great effect on the clad geometry and on dilution in the single and multi-pass aluminum matrix composite reinforced with SiC particles (Al/SiCp) coatings on ZE41 magnesium alloys deposited using a high-power diode laser (HPLD). The influence of the laser power (500-700 W), scan speed (3-17 mm/s) and laser beam focal position (focus, positive and negative defocus) on the shape factor, cladding-bead geometry, cladding-bead microstructure (including the presence of pores and cracks), and hardness has been evaluated. The correlation of these process parameters and their influence on the properties and ultimately, on the feasibility of the cladding process, is demonstrated. The importance of focal position is demonstrated. The different energy distribution of the laser beam cross section in focus plane or in positive and negative defocus plane affect on the cladding-bead properties.

  20. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  1. Oxidation Behavior of FeCrAl -coated Zirconium Cladding prepared by Laser Coating

    International Nuclear Information System (INIS)

    From the recent research trends, the ATF cladding concepts for enhanced accident tolerance are divided as follows: Mo-Zr cladding to increase the high temperature strength, cladding coating to increase the high temperature oxidation resistance, FeCrAl alloy and SiC/SiCf material to increase the oxidation resistance and strength at high temperature. To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. A laser coating method supplied with FeCrAl powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a FeCrAl-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured

  2. Silicon carbide TRIPLEX materials for CANDU fuel cladding and pressure tubes

    International Nuclear Information System (INIS)

    Ceramic Tubular Products has developed a superior silicon carbide (SiC) material TRIPLEX, which can be used for both fuel cladding and other zirconium alloy materials in light water reactor (LWR) and heavy water reactor (CANDU) systems. The fuel cladding can replace Zircaloy cladding and other zirconium based alloy materials in the reactor systems. It has the potential to provide higher fuel performance levels in currently operating natural UO2 (NEU) fuel design and in advanced fuel designs (UO2(SEU), MOX thoria) at higher burnups and power levels. In all the cases for fuel designs TRIPLEX has increased resistance to severe accident conditions. The interaction of SiC with steam and water does not produce an exothermic reaction to produce hydrogen as occurs with zirconium based alloys. In addition the absence of creep down eliminates clad ballooning during high temperature accidents which occurs with Zircaloy blocking water channels required to cool the fuel. (author)

  3. The state-of-the-art laser bio-cladding technology

    Science.gov (United States)

    Liu, Jichang; Fuh, J. Y. H.; Lü, L.

    2010-11-01

    The current state and future trend of laser bio-cladding technology are discussed. Laser bio-cladding is used in implants including fabrication of metal scaffolds and bio-coating on the scaffolds. Scaffolds have been fabricated from stainless steel, Co-based alloy or Ti alloy using laser cladding, and new laser-deposited Ti alloys have been developed. Calcium phosphate bioceramic coatings have been deposited on scaffolds with laser to improve the wear resistence and corrosion resistence of implants and to induce bone regeneration. The types of biomaterial devices currently available in the market include replacement heart valve prosthesis, dental implants, hip/knee implants, catheters, pacemakers, oxygenators and vascular grafts. Laser bio-cladding process is attracting more and more attentions of people.

  4. PWR cladding optimization for enhanced performance margins

    International Nuclear Information System (INIS)

    As the nuclear power generating industry has matured there is an increasing trend in core operating fuel duties. This drives a continuing evolution of cladding materials, to provide performance margin and support even higher fuel duty designs. Westinghouse has developed an optimized version of ZIRLOTM, with a thin level reduced from the nominal standard ZIRLO level of 1% to a range of 0.6% to 0.8%. The lower tin level has been shown to reduce the clad corrosion of fuel rods during reactor core operation by 30% or more while still providing the mechanical and off-normal corrosion protection benefits associated with tin alloy additions. Peak oxide levels of only 20-30 μm are observed at burnups up to 63 MWd/kgU. Using relatively small changes in the final annealing temperature, the clad creep can be adjusted to meet target ranges. In-reactor measurements of creep and growth of Optimized ZIRLOTM verify mechanical characteristics equivalent to standard ZIRLO. (author)

  5. Clad ballooning model in MELCOR

    International Nuclear Information System (INIS)

    Clad ballooning may substantially decrease the flow of fluids through the affected core region and may expose the inner cladding surface to oxidation in the vicinity of rupture sites. The cladding ballooning model was not included in MELCOR 1.8.4. and consideration of incorporating the cladding ballooning model is scheduled as a post-1.8.4 release activity. The purpose of this paper is to analyze the effect of the clad ballooning model by the modified MELCOR 1.8.4 with this model. The typical accident sequence of a large LOCA scenario is selected. The clad ballooning model accelerates the accident progression compared to that without the ballooning model. The amount of hydrogen does not change much and it may be caused by ignoring the effect of flow area change. Future study is planning to analyze the flow redistribution

  6. Development of SFR Fuel Cladding Tube Materials

    International Nuclear Information System (INIS)

    A R and D program for new materials for SFR cladding tube was started in 2007. The purpose of the R and D program is to develop new cladding materials having a higher creep rupture strength than the Gr.92 steel. For this purpose, the minor alloying elements such as V, Ti, C and N were added into the ferritic/martensitic (FM) steels. 5 new alloys were designed, manufactured and evaluated. Increase of V concentration caused the increase of mass fraction of V-rich MX particles. But high V steel revealed lower yield, tensile and creep rupture strengths. High N and low C steel showed higher tensile strength and lower creep rupture strength than the low N and high C steel. The Zr addition appeared to be more effective than Ti addition in terms of yield, tensile and creep rupture strengths. In order to develop a fabrication process of SFR cladding tube, the effects of the fabrication process parameters such as a tempering temperature, cold rolling and annealing condition on the precipitates and mechanical properties of a normalized FM steel were also evaluated. Nb-rich MX precipitates were found in the specimen tempered at 550''oC while M23C6, Nb- and V-rich MX ones were observed in the specimen tempered at 750''oC. A cold rolling and an annealing at 750''oC of the specimen tempered at 550''oC induced the formation of large inhomogeneous M23C6 carbides, causing a reduced tensile strength. However, the cold rolling of the specimen tempered at 750''oC provided fine precipitates mainly due to a fragmentation of the M23C6 carbides, and an annealing at 700''oC for 30 min was found to be suitable to recover the degraded mechanical properties from a cold working. (author)

  7. Laser cladding of Al-Si/SiC composite coatings : Microstructure and abrasive wear behavior

    NARCIS (Netherlands)

    Anandkumar, R.; Almeida, A.; Vilar, R.; Ocelik, V.; De Hosson, J.Th.M.

    2007-01-01

    Surface coatings of an Al-Si-SiC composite were produced on UNS A03560 cast Al-alloy substrates by laser cladding using a mixture of powders of Al-12 wt.% Si alloy and SiC. The microstructure of the coatings depends considerably on the processing parameters. For a specific energy of 26 MJ/m2 the mic

  8. Micro structural evaluation of fuel clad chemical interaction for metallic fuels for fast reactor

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels because of higher breeding ratio and lower doubling time. Two design concepts have been proposed: one based on sodium bonded ternary alloy fuel of U-Pu-Zr ( 2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy', as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. Fuel clad chemical interaction is a serious issue limiting the life of a fuel pin as a result of formation of low temperature eutectic between the fuel and components of the cladding material. The eutectic reaction temperature between T91 and Uranium were estimated by dilatometry, differential thermal analysis and high temperature microscopy. Diffusion couple experiments were also carried out between U/Zr/T91 and U/T91 by isothermal annealing of the couples between 550 deg C to 750 deg C for times up to 1500 hrs. to find out the extent of chemical interaction. These studies were supported by metallographic examination, micro hardness measurement, XRD, SEM/EDAX and EPMA. The eutectic temperature was found to be higher than the estimated fuel clad interface temperature under the reactor operating condition. The paper highlights the results of these studies and attempts to analyze them in the light of performance. The outcome of these studies has been useful to the fuel designer in optimizing the design features and predicting the in-reactor fuel behavior. (author)

  9. Reidual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. Lower residual stresses are caused by reduced thickness of the components. As the heat input is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximately constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small. (Auth.)

  10. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  11. Improved LWR Cladding Performance by EPD Surface Modification Technique

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Sridharan, Kumar

    2012-11-26

    This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation

  12. Results of NDE Technique Evaluation of Clad Hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Dennis C. Kunerth

    2014-09-01

    This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used to detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing

  13. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: • Replaced zirconium-alloy with alternate cladding concepts: 310SS, FeCrAl, 304SS, APMT and SiC. • Performed parametric study to match reactivity lifetime requirements of Zircaloy base case. • Analyzed reactivity coefficients, spectral hardening, fission power profiles and Pu inventory. • Assessed fuel cost changes when replacing Zircaloy cladding. - Abstract: A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (∼0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical

  14. Removing hydrochloric acid exhaust products from high performance solid rocket propellant using aluminum-lithium alloy.

    Science.gov (United States)

    Terry, Brandon C; Sippel, Travis R; Pfeil, Mark A; Gunduz, I Emre; Son, Steven F

    2016-11-01

    Hydrochloric acid (HCl) pollution from perchlorate based propellants is well known for both launch site contamination, as well as the possible ozone layer depletion effects. Past efforts in developing environmentally cleaner solid propellants by scavenging the chlorine ion have focused on replacing a portion of the chorine-containing oxidant (i.e., ammonium perchlorate) with an alkali metal nitrate. The alkali metal (e.g., Li or Na) in the nitrate reacts with the chlorine ion to form an alkali metal chloride (i.e., a salt instead of HCl). While this technique can potentially reduce HCl formation, it also results in reduced ideal specific impulse (ISP). Here, we show using thermochemical calculations that using aluminum-lithium (Al-Li) alloy can reduce HCl formation by more than 95% (with lithium contents ≥15 mass%) and increase the ideal ISP by ∼7s compared to neat aluminum (using 80/20 mass% Al-Li alloy). Two solid propellants were formulated using 80/20 Al-Li alloy or neat aluminum as fuel additives. The halide scavenging effect of Al-Li propellants was verified using wet bomb combustion experiments (75.5±4.8% reduction in pH, ∝ [HCl], when compared to neat aluminum). Additionally, no measurable HCl evolution was detected using differential scanning calorimetry coupled with thermogravimetric analysis, mass spectrometry, and Fourier transform infrared absorption. PMID:27289269

  15. GCFR core cladding temperature limits

    International Nuclear Information System (INIS)

    This paper reviews the phenomena that affect selection of the GCFR cladding faulted temperature limit. The limiting effects are determined to be clad melting, strength and oxidation rate. The selected temperature limit is 13000C (23700F). The limits for normal, upset and emergency events are also breifly reviewed, and some changes under consideration are discussed

  16. Aerogel-clad optical fiber

    Science.gov (United States)

    Sprehn, Gregory A.; Hrubesh, Lawrence W.; Poco, John F.; Sandler, Pamela H.

    1997-01-01

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency.

  17. Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

    Science.gov (United States)

    Zeng, Yimin; Guzonas, David

    2016-02-01

    The supercritical water-cooled reactor (SCWR) is an innovative next generation reactor that offers many promising features, but the high-temperature high-pressure coolant introduces unique challenges to the long-term safe and reliable operation of in-core components, in particular the fuel cladding. To achieve high thermal efficiency, the Canadian SCWR concept has a coolant core outlet temperature of 625°C at 25 MPa with a peak cladding temperature as high as 800°C. International and Canadian research programs on corrosion issues in supercritical water have been conducted to support the SCWR concept. This paper provides a brief review of corrosion in supercritical water and summarizes the Canadian corrosion assessment work on potential fuel cladding materials. Five alloys, SS 347H, SS310S, Alloy 800H, Alloy 625 and Alloy 214, have been shown to have sufficient corrosion resistance to be used as the fuel cladding. Additional work, including tests in an in-reactor loop, is needed to confirm that these alloys would work as the fuel cladding in the Canadian SCWR.

  18. Formation of Hard Surfacing Layers of WC-Co with Electron Beam Cladding Method

    Science.gov (United States)

    Abe, Nobuyuki; Morimoto, Junji

    Hard surfacing layers of WC-Co/Ni-base self-fluxing alloy were successfully formed on a steel substrate using an electron beam cladding method. The WC particles were densely and homogenously dispersed within the Ni-base self-fluxing alloy without porosity. The effect of the electron beam conditions on layer formation was investigated, and the cladding layer properties were examined by hardness tests, abrasive wear tests and immersion corrosion tests. It was found that the cladding layers showed higher hardness and abrasion resistance with increasing WC-Co mixing ratio, however, corrosion resistance decreased with WC-Co mixing ratio. A coating layer having high abrasive and corrosion resistance simultaneously was achieved by multiple cladding of high WC-Co mixing ratio layers after low WC-Co mixing ratio layers.

  19. Quality assurance and quality control in fabrication of cladding tubes

    International Nuclear Information System (INIS)

    Zircaloy 2 and 4 are the most important Zirconium alloys for use as fuel cladding material in light and heavy water reactors. In fast breeder reactors the cladding tubes are of a modified 16/16 - Cr-Ni-type with improved mechanical, long - term creep rate and rupture - life versus temperature properties. Starting with hot-extruded tube shells the fabrication of Zircaloy cladding tubes is done by 3 - 4 cold reduction steps in tube reducers or rolling machines followed by heat treatments in vacuum. To obtain the specified properties a precise combination of final area reduction and final annealing is absolutely necessary. The fabrication route of stainless steel claddings and guide tubes is similar to the Zircaloy production, exceptionally the last cold-forming steps are made on cold-drawing henches, hecause of economic reasons. After each cold reduction the material is annealed at recrystalisation temperatures under protective atmosoheres. For obtaining the same final tube properties for a longer nroduction neriod the implementation of a quality assurance and control system naturally independent of the production is necessary. The application of this system regarding some of the important properties of fuel cladding tubes is reported. (RW)

  20. Temperature and humidity effects on the corrosion of aluminium-base reactor fuel cladding materials during dry storage

    International Nuclear Information System (INIS)

    The effect of temperature and relative humidity on the high temperature (up to 200 deg. C) corrosion of aluminum cladding alloys was investigated for dry storage of spent nuclear fuels. A dependency on alloy type and temperature was determined for saturated water vapor conditions. Models were developed to allow prediction of cladding behaviour of 1100, 5052, and 6061 aluminum alloys for up to 50+ years at 100% relative humidity. Calculations show that for a closed system, corrosion stops after all moisture and oxygen is used up during corrosion reactions with aluminum alloys. (author)

  1. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  2. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    International Nuclear Information System (INIS)

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  3. Screening of advanced cladding materials and UN–U3Si5 fuel

    International Nuclear Information System (INIS)

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U3Si5 fuels have the potential to exhibit reactor physics and fuel management performance similar to UO2. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN–U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN–U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels

  4. Iodine induced stress corrosion cracking of Zircaloy fuel cladding materials

    International Nuclear Information System (INIS)

    This report documents the work performed by the Co-ordinated Research Project (CRP) on Stress Corrosion Cracking of Zirconium Alloy Fuel Cladding. The project consisted of out-of-pile laboratory measurements of crack propagation rates in Zircaloy sheet specimens in an iodine containing atmosphere. The project was overseen by a supervisory group consisting of experts in the field, who also contributed a state of the art review. This report describes all of the work undertaken as part of the CRP, and includes: a review of the state of the art understanding of stress corrosion cracking behaviour of zirconium alloy cladding material; a description of the experimental equipment, test procedures, material characterizations and test matrix; discussion of the work undertaken by the host laboratory and the specific contributions by each of the four participant laboratories; a compilation of all experimental results obtained; and the supervisory group's analysis and discussion of the results, plus conclusions and recommendations

  5. Corrosion of spent nuclear fuel aluminium cladding in ordinary water

    International Nuclear Information System (INIS)

    Corrosion of aluminium alloy cladding of spent nuclear fuel elements in ordinary water is examined in the spent fuel storage pool of the RA research reactor at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro. Experimental examinations are carried out within framework of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water', Phase II. Racks with coupons made of different aluminium alloys were exposed to water influence for period of six months to six years. The project comprises also activities on monitoring of the water chemistry and radioactivity in the storage pool. Visual and microscopic examinations of surfaces of aluminium coupons of the test racks have been done recently and results were presented in this paper confirming strong influence of water quality and exposition time to corrosion process. (author)

  6. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  7. Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

    International Nuclear Information System (INIS)

    Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs

  8. Microstructure and wear-resistance of laser clad TiC particle-reinforced coating

    NARCIS (Netherlands)

    Lei, T.C.; Ouyang, J.H.; Pei, Y.T.; Zhou, Y.

    1995-01-01

    A TiC-Ni alloy composite coating was clad to 1045 steel substrate using a 2kW CO2 laser. The microstructural constituents of the clad layer are found to be gamma-Ni and TiCp in the dendrites, and a fine eutectic of gamma-Ni plus (Fe, Cr)(23)C-6 in the interdendritic areas. Partial dissolution and ag

  9. Investigations into the cladding of nuclear materials using the plasma hot wire process

    International Nuclear Information System (INIS)

    Investigations of the fusion weld cladding of 22NiMoCr37 and 20MnMoNi55 steels by an austenitic 18/8 steel and a nickel base chromium alloy are described. Metallographic, intercrystalline corrosion, bend, tensile, notch impact and underclad cracking tests were carried out. Results indicate that the PHC process can be considered as a significant complement to existing fusion weld cladding processes. (U.K.)

  10. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  11. Modelling nuclear fuel behaviour and cladding viscoelastic response

    OpenAIRE

    Tulkki, Ville

    2015-01-01

    In light water reactors the nuclear fuel is in the form of uranium dioxide pellets stacked inside a thin-walled tube made from Zirconium alloy. The fuel rods provide the first barriers to the release of radioactivity as the isotopes are contained within the fuel matrix and the cladding tubes. Fuel behaviour analysis investigates the state of the fuel at given boundary conditions and irradiation history. The scope of this thesis consists of two main themes. The first is the uncertainty and ...

  12. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  13. Microstructure and formation of melting zone in the interface of Ti/NiCr explosive cladding bar

    International Nuclear Information System (INIS)

    Highlights: ► CP-Ti/NiCr alloy bar used in medical treatment was made by explosive cladding. ► Local melting zones are encountered in sections of the cladding interface. ► Melting zone is composed of intermetallics of brittle and without element diffusion. ► High solidification rate is the major reason for the formation of the melting zone. - Abstract: The tube of titanium and the bar of NiCr alloy were bonded through explosive cladding technique; a good quality bonding was obtained. Melting zones are encountered in sections of the explosive cladding interface, and they serious affect the properties of the explosive cladding composite. Microstructure of melting zone in the interface of Ti/NiCr explosive cladding bar were investigated by means of optical microscope (OM), scanning electron microscope (SEM) and microhardness as well as using micro-focus X-ray diffraction and electron probe analyses. The results show that the melting zone is composed of intermetallics of brittle and stiff, and there is no element diffusion during explosive cladding process. The tendency of composition segregation of melting zone is decreased. The solidification rate and the actual distribution coefficient of solute in the melting zone of Ti/NiCr explosive cladding interface are not less than 0.1 × 108 k/s and 1 respectively. The formation of microstructure in the melting zone is result from the high solidification rate in the explosive cladding interface

  14. Hydrogen interactions in aluminum-lithium alloys

    Science.gov (United States)

    Smith, S. W.; Scully, J. R.

    1991-01-01

    A program is described which seeks to develop an understanding of the effects of dissolved and trapped hydrogen on the mechanical properties of selected Al-Li-Cu-X alloys. A proposal is made to distinguish hydrogen (H2) induced EAC from aqueous dissolution controlled EAC, to correlate H2 induced EAC with mobile and trapped concentrations, and to identify significant trap sites and hydride phases (if any) through use of model alloys and phases. A literature review shows three experimental factors which have impeded progress in the area of H2 EAC for this class of alloys. These are as listed: (1) inter-subgranular fracture in Al-Li alloys when tested in the S-T orientation in air or vacuum make it difficult to readily detect H2 induced fracture based on straight forward changes in fractography; (2) the inherently low H2 diffusivity and solubility in Al alloys is further compounded by a native oxide which acts as a H2 permeation barrier; and (3) H2 effects are masked by dissolution assisted processes when mechanical testing is performed in aqueous solutions.

  15. Electrochemical profiling of multi-clad aluminium sheets used in automotive heat exchangers

    DEFF Research Database (Denmark)

    Bordo, Kirill; Ambat, Rajan; Peguet, Lionel;

    2014-01-01

    The objective of the present study is to understand the mechanisms of corrosion propagation across the multi-clad structure of Al alloys sheets as a function of local alloy composition and microstructure, with and without brazing treatment. Electro-chemical behaviour at different depths...... was profiled using a combination of glow dis-charge optical emission spectroscopy (GDOES) sputtering, localized potentiodynam-ic polarization and zero resistance ammetry (ZRA) measurements. Multi-clad struc-ture used was a four layer sandwich consisting of a copper-containing AA3xxx long-life core alloy, AA......4343 brazing clad on both sides and a copper-free AA3xxx interlay-er on the air-side of the sandwich sheet. The polarization behaviour of both as-rolled and brazed materials (i.e. corrosion potential, pitting potential, cathodic and anodic reactivities) was determined as a function of depth using...

  16. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  17. Simulation and Analysis on Hoop Strength Test of Multilayered SiC Composite Fuel Cladding

    International Nuclear Information System (INIS)

    Silicon carbide-based ceramics and their composites have been studied for fusion and advanced fission energy systems. For fission reactors, SiCf/SiC composites can be applied to core structural materials. Multi-layered SiC composite fuel cladding, which consists of monolith inner/outer layer and intermediate SiCf/SiC composite layer is one of candidates for a replacement for the zirconium alloy cladding, owing to the superior high temperature strength and low hydrogen generation under severe accident conditions. The SiC composite cladding has to retain the mechanical properties and its structure from the inner pressure caused by fission products to apply a cladding of fission reactor. The inner pressure caused by fission products induces hoop stress in a circumferential direction. Hoop strength test using expandable polyurethane plug is designed for evaluating the mechanical properties of fuel cladding. In this paper, hoop strength test of the multilayered cladding was simulated in order to evaluate hoop stress and shear stress at the cladding and the fracture of the cladding was analyzed

  18. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    International Nuclear Information System (INIS)

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings

  19. Irradiation Performance of Oxide Dispersion Strengthened (ODS) Ferritic Steel Claddings for Fast Reactor Fuels

    International Nuclear Information System (INIS)

    Irradiation tests in Joyo and BOR-60 for the ODS claddings developed by JAEA were carried out in order to confirm the irradiation performance of the ODS claddings and thus judge their applicability to high burnup and high temperature fast reactor fuels. The main points of the tests are summarized as follows. 1) Valuable data indicating application prospects of the ODS claddings for high burnup fuels were obtained regarding superior dimensional stability and integrity of the upper end-plug welded by the PRW method. 2) No significant irradiation effect on mechanical properties of the ODS claddings was confirmed within the irradiation conditions in the Joyo material irradiation tests. The oxide particles and microstructures of ODS claddings were confirmed to be stable during neutron irradiation. 3) FCCI data for the ODS claddings were acquired within the irradiation conditions in the BOR-60 fuel pin irradiation tests, and it was shown that FCCI could be reduced by lowering oxygen potential in the fuel, even for low Cr content claddings such as 9Cr-ODS steel. 4) The manufacturing technology development applied to the full pre-alloy process to improve homogeneity of the ODS cladding has already started, and the expected results are being obtained

  20. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The hot and cold cracking tendencies of some early iron aluminide alloy compositions have limited their use in applications where good weldability is required. Using hot crack testing techniques invented at ORNL, and experimental determinations of preheat and postweld heat treatment needed to avoid cold cracking, we have developed iron aluminide filler metal compositions which can be successfully used to weld overlay clad various substrate materials, including 9Cr-1Mo steel, 2-1/4Cr-1Mo steel, and 300-series austenitic stainless steels. Dilution must be carefully controlled to avoid crack-sensitive deposit compositions. The technique used to produce the current filler metal compositions is aspiration-casting, i.e. drawing the liquid from the melt into glass rods. Future development efforts will involve fabrication of composite wires of similar compositions to permit mechanized gas tungsten arc (GTA) and/or gas metal arc (GMA) welding.

  1. Creep anisotropy of Zircaloy cladding tubes

    International Nuclear Information System (INIS)

    First of all, a survey is given on the texture of Zircaloy cladding tubes obtained depending on the manufacturing conditions, and the state of knowledge on the anisotropy of the mechanical properties of the zirconium alloys connected with the texture is outlined. Theoretical formulations are set up for the phenomenological representation of the anisotropic creep. The results of tension and compression tests and the thus obtained creep site curves exhibit distinct differences with tubes having different textures. Furthermore, on asymmetry regarding compressive tensile stress is found in such a manner that the material under compression stress is more resistant to creep. Finally, discussions follow on the deformation mechanisms and a comparison with flow processes as well as indications on the significance of these creep results within the framework of fuel rod design are given. (IHoe/LH)

  2. Factors controlling hydrogen cracking during cladding of nuclear vessel steels

    International Nuclear Information System (INIS)

    During cladding of low alloy steels in nuclear pressure vessels for corrosion resistance, a potential problem exists of underclad hydrogen cracking. Research was undertaken to gain a better insight into the factors controlling underclad hydrogen cracking during cladding A508 Cl 3 nuclear vessel steels and to ensure the continued development of safe welding procedures in this critical application. The project was divided into three experimental phases. Phase I studied the potential and deposit hydrogen levels in Type 309 austenitic stainless steel and Ni alloy consumables and weld metals. Phase II incorporated implant testing of the A508 Cl 3 base material. A large test panel was fabricated in Phase III to approach the conditions of restraint and heat sink that are present in the pressure vessel cladding operation, but not necessarily those of the most critical components, such as nozzles where the cylindrical geometry may increase the overall restraint. The A508 Cl 3 test material was electron beam welded into the center of the test block which was then submerged arc-strip clad using very severe welding conditions in an attempt to generate underclad hydrogen cracks. It was found that for the shielded metal-arc welding (SMAW) and submerged arc welding (SAW) processes, deposit hydrogen levels were primarily controlled by flux moisture content. With single layer deposition, the implant test did not show evidence of the influence of segregation on cold cracking. All SMAW implant tests, without preheat and regardless of consumable, gave lower critical stress thresholds below about 51 ksi. A preheat of 150 deg.C increased this threshold to 80 ksi with Type 306 consumables. Even under welding conditions favorable for cracking, underclad hydrogen cracks could not be developed in a large-scale simulation of a cladding operation, indicating that very high total system restraint is needed to induce cracking

  3. Novel strip-cast Mg/Al clad sheets with excellent tensile and interfacial bonding properties

    Science.gov (United States)

    Kim, Jung-Su; Lee, Dong Ho; Jung, Seung-Pill; Lee, Kwang Seok; Kim, Ki Jong; Kim, Hyoung Seop; Lee, Byeong-Joo; Chang, Young Won; Yuh, Junhan; Lee, Sunghak

    2016-06-01

    In order to broaden industrial applications of Mg alloys, as lightest-weight metal alloys in practical uses, many efforts have been dedicated to manufacture various clad sheets which can complement inherent shortcomings of Mg alloys. Here, we present a new fabrication method of Mg/Al clad sheets by bonding thin Al alloy sheet on to Mg alloy melt during strip casting. In the as-strip-cast Mg/Al clad sheet, homogeneously distributed equi-axed dendrites existed in the Mg alloy side, and two types of thin reaction layers, i.e., γ (Mg17Al12) and β (Mg2Al3) phases, were formed along the Mg/Al interface. After post-treatments (homogenization, warm rolling, and annealing), the interfacial layers were deformed in a sawtooth shape by forming deformation bands in the Mg alloy and interfacial layers, which favorably led to dramatic improvement in tensile and interfacial bonding properties. This work presents new applications to multi-functional lightweight alloy sheets requiring excellent formability, surface quality, and corrosion resistance as well as tensile and interfacial bonding properties.

  4. Novel strip-cast Mg/Al clad sheets with excellent tensile and interfacial bonding properties.

    Science.gov (United States)

    Kim, Jung-Su; Lee, Dong Ho; Jung, Seung-Pill; Lee, Kwang Seok; Kim, Ki Jong; Kim, Hyoung Seop; Lee, Byeong-Joo; Chang, Young Won; Yuh, Junhan; Lee, Sunghak

    2016-01-01

    In order to broaden industrial applications of Mg alloys, as lightest-weight metal alloys in practical uses, many efforts have been dedicated to manufacture various clad sheets which can complement inherent shortcomings of Mg alloys. Here, we present a new fabrication method of Mg/Al clad sheets by bonding thin Al alloy sheet on to Mg alloy melt during strip casting. In the as-strip-cast Mg/Al clad sheet, homogeneously distributed equi-axed dendrites existed in the Mg alloy side, and two types of thin reaction layers, i.e., γ (Mg17Al12) and β (Mg2Al3) phases, were formed along the Mg/Al interface. After post-treatments (homogenization, warm rolling, and annealing), the interfacial layers were deformed in a sawtooth shape by forming deformation bands in the Mg alloy and interfacial layers, which favorably led to dramatic improvement in tensile and interfacial bonding properties. This work presents new applications to multi-functional lightweight alloy sheets requiring excellent formability, surface quality, and corrosion resistance as well as tensile and interfacial bonding properties. PMID:27245687

  5. High power cladding light strippers

    Science.gov (United States)

    Wetter, Alexandre; Faucher, Mathieu; Sévigny, Benoit

    2008-02-01

    The ability to strip cladding light from double clad fiber (DCF) fibers is required for many different reasons, one example is to strip unwanted cladding light in fiber lasers and amplifiers. When removing residual pump light for example, this light is characterized by a large numerical aperture distribution and can reach power levels into the hundreds of watts. By locally changing the numerical aperture (N.A.) of the light to be stripped, it is possible to achieve significant attenuation even for the low N.A. rays such as escaped core modes in the same device. In order to test the power-handling capability of this device, one hundred watts of pump and signal light is launched from a tapered fusedbundle (TFB) 6+1x1 combiner into a high power-cladding stripper. In this case, the fiber used in the cladding stripper and the output fiber of the TFB was a 20/400 0.06/0.46 N.A. double clad fiber. Attenuation of over 20dB in the cladding was measured without signal loss. By spreading out the heat load generated by the unwanted light that is stripped, the package remained safely below the maximum operating temperature internally and externally. This is achieved by uniformly stripping the energy along the length of the fiber within the stripper. Different adhesive and heat sinking techniques are used to achieve this uniform removal of the light. This suggests that these cladding strippers can be used to strip hundreds of watts of light in high power fiber lasers and amplifiers.

  6. The influence of copper on Zircaloy spent fuel cladding degradation under a potential tuff repository condition

    International Nuclear Information System (INIS)

    This paper reports the results of an experiment designed to detect the influence of copper on Zircaloy spent fuel cladding degradation in one possible repository environment. Copper and copper alloys are being considered for use in a tuff repository. The compatibility of a copper waste package container and the Zircaloy cladding on spent fuel has been questioned essentially because copper ion has been observed to accelerate zirconium alloy corrosion in acid environments, as does ferric iron, and a phenomenon called ''crud-induced localized corrosion'' is observed in some Boiling Water Reactors where thorugh-the-wall corrosion pits develop beneath copper-rich crud deposits. 16 refs., 6 figs., 2 tabs

  7. Cladding properties changes during operation

    International Nuclear Information System (INIS)

    Austenitic cladding was originally designed as a protection of ferritic/bainitic base materials of reactor pressure vessels against corrosion. Nevertheless, its existence must be taken into account into reactor pressure vessel integrity evaluation from several reasons: cladding has different thermal properties with respect to base metal which affect temperature fields in a vessel; cladding has different mechanical and thermal-mechanical properties comparing with base metal which affect stress field in a vessel; austenitic cladding has different fracture mechanics properties that base metal, but they are also changing during operation due to radiation damage. Austenitic cladding from WWER-440 reactor pressure vessels has been studied within an extended surveillance programme and some interesting results have been obtained. Austenitic cladding made from Nb-stabilized 18/10 type is characterized by some δ-ferrite content in its initial state which results in slight transition behaviour of fracture properties. These properties are changing after irradiation - fracture toughness is decreasing as well as tensile properties are increasing. This second trend was also supported by measurements realized during in-service inspections of inner vessel wall using instrumented indentation testing method. Knowledge of austenitic properties, mainly of its fracture mechanics parameters, is also necessary for a proper evaluation of reactor pressure vessel behaviour during PTS regimes. (author)

  8. Advanced powder metallurgy aluminum alloys via rapid solidification technology, phase 2

    Science.gov (United States)

    Ray, Ranjan; Jha, Sunil C.

    1987-01-01

    Marko's rapid solidification technology was applied to processing high strength aluminum alloys. Four classes of alloys, namely, Al-Li based (class 1), 2124 type (class 2), high temperature Al-Fe-Mo (class 3), and PM X7091 type (class 4) alloy, were produced as melt-spun ribbons. The ribbons were pulverized, cold compacted, hot-degassed, and consolidated through single or double stage extrusion. The mechanical properties of all four classes of alloys were measured at room and elevated temperatures and their microstructures were investigated optically and through electron microscopy. The microstructure of class 1 Al-Li-Mg alloy was predominantly unrecrystallized due to Zr addition. Yield strengths to the order of 50 Ksi were obtained, but tensile elongation in most cases remained below 2 percent. The class 2 alloys were modified composition of 2124 aluminum alloy, through addition of 0.6 weight percent Zr and 1 weight percent Ni. Nickel addition gave rise to a fine dispersion of intermetallic particles resisting coarsening during elevated temperature exposure. The class 2 alloy showed good combination of tensile strength and ductility and retained high strength after 1000 hour exposure at 177 C. The class 3 Al-Fe-Mo alloy showed high strength and good ductility both at room and high temperatures. The yield and tensile strength of class 4 alloy exceeded those of the commercial 7075 aluminum alloy.

  9. Thermodynamic Database for Zirconium Alloys

    OpenAIRE

    Jerlerud Pérez, Rosa

    2006-01-01

    For many decades zirconium alloys have been commonly used in the nuclear power industry as fuel cladding material. Besides their good corrosion resistance and acceptable mechanical properties the main reason for using these alloys is the low neutron absorption. Zirconium alloys are exposed to a very severe environment during the nuclear fission process and there is a demand for better design of this material. To meet this requirement a thermodynamic database is useful to support material desi...

  10. Nuclear fuel cladding tube and method of manufacturing the same

    International Nuclear Information System (INIS)

    A cladding tube main body made of a zirconium alloy and an end plug are joined by welding. Tensile stresses at the weld heat-affected portion between the cladding tube main body and the end plug are removed, so that compression stresses of 0 MPa or more but less than the endurance strength of the zirconium alloy is applied on the weld heat affected portion. As the zirconium alloy, a zircaloy-2 or zircaloy-4 is preferable since it is excellent in the corrosion resistance and strength. The zirconium alloy may preferably be used also to the material of the end plug. The treatment for the removal of the tensile stresses includes a method of applying annealing to the weld heat-affected portion or a method of applying compression stresses thereto by applying external force such as a shot peening treatment. This can suppress occurrence of nodular corrosion and white homogeneous corrosion caused in the vicinity of the welded portion. (I.N.)

  11. Optimization of Hydride Rim Formation in Unirradiated Zr 4 Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Hanson, Brady D.; MacFarlan, Paul J.

    2013-09-30

    The purpose of this work is to build on the results reported in the M2 milestone M2FT 13PN0805051, document number FCRD-USED-2013-000151 (Hanson, 2013). In that work, it was demonstrated that unirradiated samples of zircaloy-4 cladding could be pre-hydrided at temperatures below 400°C in pure hydrogen gas and that the growth of hydrides on the surface could be controlled by changing the surface condition of the samples and form a desired hydride rim on the outside diameter of the cladding. The work performed at Pacific Northwest National Laboratory since the issuing of the M2 milestone has focused its efforts to optimize the formation of a hydride rim on available zircaloy-4 cladding samples by controlling temperature variation and gas flow control during pre-hydriding treatments. Surface conditioning of the outside surface was also examined as a variable. The results of test indicate that much of the variability in the hydride thickness is due to temperature variation occurring in the furnaces as well as how hydrogen gas flows across the sample surface. Efforts to examine other alloys, gas concentrations, and different surface conditioning plan to be pursed in the next FY as more cladding samples become available

  12. Production and inspection of zircaloy fuel cladding tubes

    International Nuclear Information System (INIS)

    Zircaloy fuel cladding tubes are used for light and heavy water reactors. The tubes are basically produced in accordance with the ASTM B353 ''Standard specification for wrought zirconium and zirconium alloy seamless and welded tubes for nuclear service''. The production procedure for the zircaloy tubes is composed of consumable electrode are melting, forging, heat treatment, extruding, cold rolling, annealing, final roll reform and surface grinding. Concerning these producing procedure, the key points of each process relating to the material characteristics and the producing machines are presented. Next, the inspection of zircaloy fuel cladding tubes is outlined. The inspection standard of ASTM B 353-77 is tabulated as an example. Ultrasonic inspection and surface visual inspection as the flaw inspection methods, the dimensional inspection by ultrasonic pulse method for measuring the diameter and the wall thickness, electric and air micrometers for measuring the inner and outer diameters, and the ultrasonic resonance method for measuring the wall thickness, and the straightness inspection of tubes using a surface plate are explained. The mechanical tests for the zircaloy cladding tubes, such as the tensile test and the burst test, are described. The metal structure test, the corrosion test and the chemical analysis are outlined, and the characteristics of zircaloy cladding tubes for BWRs and PWRs are tabulated. (Nakai, Y.)

  13. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    International Nuclear Information System (INIS)

    The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 5700C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 2700C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 3800C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 4000C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

  14. Development of advanced cladding material for burnup extension

    International Nuclear Information System (INIS)

    The development of new cladding materials is one of the critical issues on burnup extension. The practical life of Zircaloy would be limited by the growth of oxide films and by the ductility loss due to hydride precipitation, oxygen absorption and radiation damage. In the case of high burnup using MOX fuels, the low neutron adsorption cross section of Zircaloy is not a dominant factor for selecting the cladding material, because MOX fuels can be enriched up to 20%Pu. Austenitic stainless steel, titanium alloy, niobium alloy, ferritic steel and nickel base superalloy are considered as candidate materials. The corrosion resistance, mechanical properties and the irradiation resistance of these materials were examined for evaluating the practical possibility as a cladding material. The austenitic stainless steel with high g phase stability was selected as the primary candidate material. However, it is required to improve the resistance to irradiation associated stress corrosion cracking through the experience in LWR plants. In the JAERI, the austenitic stainless steel with intergranular corrosion resistance has been developed by the adjustment of the chemical composition, the modification of the metallographic structure by thermo-mechanical treatment and the purification by electron beam melting. (author)

  15. Optimization of Ni-Based WC/Co/Cr Composite Coatings Produced by Multilayer Laser Cladding

    OpenAIRE

    Andrea Angelastro; Sabina L. Campanelli; Giuseppe Casalino; Antonio D. Ludovico

    2013-01-01

    As a surface coating technique, laser cladding (LC) has been developed for improving wear, corrosion, and fatigue properties of mechanical components. The main advantage of this process is the capability of introducing hard particles such as SiC, TiC, and WC as reinforcements in the metallic matrix such as Ni-based alloy, Co-based alloy, and Fe-based alloy to form ceramic-metal composite coatings, which have very high hardness and good wear resistance. In this paper, Ni-based alloy (Colmonoy ...

  16. Development and Processing Improvement of Aerospace Aluminum Alloys

    Science.gov (United States)

    Lisagor, W. Barry; Bales, Thomas T.

    2007-01-01

    This final report, in multiple presentation format, describes a comprehensive multi-tasked contract study to improve the overall property response of selected aerospace alloys, explore further a newly-developed and registered alloy, and correlate the processing, metallurgical structure, and subsequent properties achieved with particular emphasis on the crystallographic orientation texture developed. Modifications to plate processing, specifically hot rolling practices, were evaluated for Al-Li alloys 2195 and 2297, for the recently registered Al-Cu-Ag alloy, 2139, and for the Al-Zn-Mg-Cu alloy, 7050. For all of the alloys evaluated, the processing modifications resulted in significant improvements in mechanical properties. Analyses also resulted in an enhanced understanding of the correlation of processing, crystallographic texture, and mechanical properties.

  17. Effect of Niobium on the Microstructure and Wear Resistance of Nickel-Based Alloy Coating by Laser Cladding%Nb对激光熔覆镍基合金涂层显微组织和磨损性能的影响

    Institute of Scientific and Technical Information of China (English)

    董刚; 严彪; 邓琦林; 余廷

    2011-01-01

    利用激光熔覆在45#钢表面制备不同铌含量的镍基合金复合涂层.使用扫描电镜(SEM)、X射线衍射(XRD)仪等仪器分析涂层横断面的显微组织.结果表明:铌改性的镍基合金复合涂层不仅含有γ-Ni树枝晶、枝晶间的共晶组织、CrB型硼化物,还含有大量的弥散分布NbC颗粒.Nb元素的添加,使镍基合金复合涂层中的碳化物以NbC颗粒和MC型碳化物形式析出,抑制了纯镍基合金涂层中粗大的MC型碳化物的大量析出.复合涂层的显微结构和相的转变改善了其耐磨损性能,并且镍基合金复合涂层的显微硬度和耐磨损性能随着铌含量的提高而提高.%The nickel-based alloys with different Nb contents were deposited on AISI 1045 carbon steel by laser cladding. The effect of Nb on the microstructures of the nickel-based alloy coatings was investigated using scanning electron microscopy (SEM) and X-ray diffraction (XRD) techniques. The result show that the microstructures of the Nb-modified nickel-based alloy coatings are mainly composed of γ-Ni dendrites, interdendritic eutectics, CrB type chromium borides, and dispersed NbC particles. It is found that the addition of Nb will lead to the precipitation of the NbC particles and MC type carbides instead of the MC, and MC type carbides can be observed in the Nb-free nickel-based alloy coating. The microhardness and wear resistance of the coatings increase with the increase of Nb contents. The improvement of the wear resistance of the Nb-modifled nickel-based alloy coatings is attributed to the microstructural change and phase variation.

  18. A preliminary study of laser cladding of AISI 316 stainless steel using preplaced NiTi wire

    International Nuclear Information System (INIS)

    NiTi wire of diameter 1 mm was preplaced on AISI 316 stainless steel samples by using a binder. Melting of the NiTi wire to form a clad track on the steel substrate was achieved by means of a high-power CW Nd:YAG laser using different processing parameters. The geometry and microstructure of the clad deposit were studied by optical microscopy and scanning electron microscopy (SEM), respectively. The hardness and compositional profiles along the depth of the deposit were acquired by microhardness testing and energy-dispersive spectroscopy (EDS), respectively. The elastic behavior of the deposit was analyzed using nanoindentation, and compared with that of the NiTi wire. The dilution of the NiTi clad by the substrate material beneath was substantial in single clad tracks, but could be successively reduced in multiple clad layers. A strong fusion bonding with tough interface could be obtained as evidenced by the integrity of Vickers indentations in the interfacial region. In comparison with the NiTi cladding on AISI 316 using the tungsten inert gas (TIG) process, the laser process was capable of producing a much less defective cladding with a more homogeneous microstructure, which is an essential cladding quality with respect to cavitation erosion and corrosion resistance. Thus, the present preliminary study shows that laser cladding using preplaced wire is a feasible method to obtain a thick and homogeneous NiTi-based alloy layer on AISI 316 stainless steel substrate

  19. A preliminary study of laser cladding of AISI 316 stainless steel using preplaced NiTi wire

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, F.T.; Lo, K.H.; Man, H.C

    2004-08-25

    NiTi wire of diameter 1 mm was preplaced on AISI 316 stainless steel samples by using a binder. Melting of the NiTi wire to form a clad track on the steel substrate was achieved by means of a high-power CW Nd:YAG laser using different processing parameters. The geometry and microstructure of the clad deposit were studied by optical microscopy and scanning electron microscopy (SEM), respectively. The hardness and compositional profiles along the depth of the deposit were acquired by microhardness testing and energy-dispersive spectroscopy (EDS), respectively. The elastic behavior of the deposit was analyzed using nanoindentation, and compared with that of the NiTi wire. The dilution of the NiTi clad by the substrate material beneath was substantial in single clad tracks, but could be successively reduced in multiple clad layers. A strong fusion bonding with tough interface could be obtained as evidenced by the integrity of Vickers indentations in the interfacial region. In comparison with the NiTi cladding on AISI 316 using the tungsten inert gas (TIG) process, the laser process was capable of producing a much less defective cladding with a more homogeneous microstructure, which is an essential cladding quality with respect to cavitation erosion and corrosion resistance. Thus, the present preliminary study shows that laser cladding using preplaced wire is a feasible method to obtain a thick and homogeneous NiTi-based alloy layer on AISI 316 stainless steel substrate.

  20. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  1. Clad Degradation - FEPs Screening Arguments

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2004-03-17

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796]).

  2. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  3. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  4. Laser Clad Nickel Based Superalloys: Microstructure Evolution And High Temperature Oxidation Studies

    Science.gov (United States)

    Sircar, S.; Ribaudo, C.; Mazumder, J.

    1988-10-01

    Application of alloy coatings with superior oxidation resistance at elevated temperatures (1200°C) on superalloy components is of interest at present. There is a general consensus that the addition of rare earths such as hafnium (Hf) to these alloys has a pronounced effect on their performance. An in situ laser cladding technique was used to produce Ni-Al-Cr-Hf alloys on a nickel alloy substrate. Scanning Electron Microscope (SEM), Transmission Electron Microscope (TEM), and Scanning Transmission Electron Microscope (STEM) attached with Energy Dispersive X-ray (EDX) analyzers were employed for microstructural evolution studies of alloys produced during the laser cladding process. The microstructure of these alloys mainly consists of dendrites of Y' of the Ni3Al type with about 11-14 wt% Hf and an interdendritic eutectic phase. Electron microscopy in the dendritic zones reveals ordered domains whose morphology depends on laser cladding process parameters. Variation in these parameters produced only subtle changes in the composition and cell spacing of the dendritic phase. The eutectic constituent consists of a Hf-rich phase and a Hf-lean phase in an alternating lamellar structure. Convergent beam diffraction and x-ray spectroscopy techniques were used to characterize the constituents. A possible phase transformation sequence has been suggested. Differential Thermal Analysis (DTA) work indicates that the Y' dissolution temperature for the claddings is at least as high as the substrate material (Rene 80). Single cycle oxidation tests of eight hours at 1200°C in slowly flowing air reveal that the claddings have a lower weight gain rate than the substrate itself. Microchemistry and microstructure of the oxidized samples are examined using SEM attached with EDX and Auger Electron Spectroscopic (AES) techniques. The improvement in the oxidation resistance is believed to be at least partially due to the mechanical pegging between alumina coated hafnia protrusions and the

  5. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    International Nuclear Information System (INIS)

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  6. Evaluation of fuel-cladding properties at high temperatures: Final Report, April 1988

    International Nuclear Information System (INIS)

    Current performance capabilities of fuel cladding could possibly be extended under severe accident conditions by using materials other than the Zircaloys or with clad coatings. Such materials would enhance LWR safety as they would extend the time to failure, and reduce the amount of hydrogen production and the extent of fuel clad ballooning. This study serves as a preliminaty screening tool to determine which materials would provide the desired improvements in stream oxidation resistance and elevated temperature strength in the case of severe accident conditions. The screening results showed that molybdenum and some of its alloys, and some niobium alloys, all commercially produced, could be possible candidates. The oxidation resistance of these materials as well as the Zircaloys could be enhanced when used with commercially available surface coatings. These fall into the general categories of silicides and aluminides. For the Zircaloys in particular, research would be required to develop successful coating and bonding techniques

  7. 不锈钢铝合金半固态连接工艺参数研究%Technological parameters of stainless steel-aluminum alloy semisolid joining clad

    Institute of Scientific and Technical Information of China (English)

    刘洪伟; 郭成

    2007-01-01

    By using semisolid joining technique, the bonding of stainless steel and semisolid aluminum alloy is successfully realized. The relationships between interfacial shear strength and solid fraction of aluminum alloy, bonding pressure and time of keeping pressure were studied by the method of orthogonal experiment. The interfacial structure and the fracture structure of the bonding plate are studied by means of optical microscope (OM) and scanning electron microscope (SEM). The results show that there is the best solid fraction between the solid phase line and the liquid phase line of the semisolid aluminum alloy, with the increase of bonding pressure and pressure time, the interfacial shear strength increases rapidly, and then with further increase of bonding pressure and pressure time, the shear strength rises little. Along the interface, solid phase and liquid phase bond with stainless steel by turns because of the different diffusion ability. So, a new type of non-equilibrium diffusion interfacial structure is constructed at the interface of stainless steel and aluminum alloy, compound mechanism of plastic and brittle fracture interface was formed at the shear fracture interface.

  8. Friction surface cladding: An exploratory study of a new solid state cladding process

    NARCIS (Netherlands)

    Liu, S.J.; Bor, T.C.; Stelt, van der A.A.; Geijselaers, H.J.M.; Kwakernaak, C.; Kooijman, A.M.; Mol, J.M.C.; Akkerman, R.; Boogaard, van den A.H.

    2015-01-01

    Friction surface cladding is a newly developed solid state cladding process to manufacture thin metallic layers on a substrate. In this study the influence of process conditions on the clad layer appearance and the mechanical properties of both the clad layer and the substrate were investigated. Thi

  9. Microstructure and abrasive wear studies of laser clad Al-Si/SiC composite coatings

    NARCIS (Netherlands)

    Anandkumar, R.; Colaco, R.; Ocelik, V.; De Hosson, J. Th. M.; Vilar, R.; Gyulai, J; Szabo, PJ

    2007-01-01

    Surface coatings of Al-Si/SiC metal-matrix composites were deposited on Al-7 wt. % Si alloy substrates by laser cladding. The microstructure of the coatings was characterized by optical microscopy, scanning electron microscopy (SEM) and X-ray diffraction (XRD). The microstructure of the coating mate

  10. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    International Nuclear Information System (INIS)

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI

  11. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  12. Development of high performance cladding

    International Nuclear Information System (INIS)

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  13. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  14. Clad plates for construction of apparatus

    International Nuclear Information System (INIS)

    Importance of clad plates on the field of the construction of apparatus for the chemistry and petrol chemistry. Description of a cladding process to bond permanently and integrally ferritic steels and corrosion resistant and heat resistant materials by rolling. Information on available combinations of materials and gauge as well as on indispensable requirements to be met by the quality of the material. Results of tests carried out on the bond. Distribution of the elements between the clad and the base material. Bond properties, corrosion behaviour, toughness values and tensile properties of clad plates, heat treatment, cutting and welding of clad plates. Demonstration of applications. (orig.)

  15. Fuel cladding tubes and manufacture thereof

    International Nuclear Information System (INIS)

    Purpose: To enable smooth contaction between fuel pellets and cladding tubes, as well as prevent chemical reaction for the fission products released from the pellets. Method: The inner surface of a cladding tube is coated with a copper film and further provided thereover with a graphite film. The graphite film is formed through electrophoretic coating as follows: A cladding tube is supported rotatably in an electrophoretic coating tank containing coating solution incorporated with graphite powder and connected to an anode. A cathode is attached to the inside of the cladding tube. Coating current is supplied while rotating the cladding tube and the graphite film is formed through electrophoresis. (Ikeda, J.)

  16. Zirconium fuel cladding corrosion prediction in fuel assembly operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Berezina, I.G., E-mail: kritsky@givnipiet.spb.ru, E-mail: alemaskina@givnipiet.ru [Leading Inst. ' VNIPIET' , Saint Petersburg (Russian Federation)

    2010-07-01

    At present, the work to extend fuel cycles is carried out at NPP with VVER reactors. With the increase of fuel assembly burn-up to 70-100 MWd/kg U and linear power, the local coolant «nucleate boiling» is inevitable which in combination with coolant «acidification» alongside with the existing water chemistry norms will increase zirconium alloy corrosion. The rate of Zr alloy corrosion under reactor irradiation depends on temperature and heat flux through fuel cladding, coolant chemistry (concentrations of H{sub 2}O{sub 2}, OH{sup -}, O{sub 2}, hydrogen, ammonia, strong alkalis - LiOH, KOH, pH, ets.), steam content, alloy composition and some other parameters. A generalized model for calculating Zr alloys corrosion, which take into account the above-mentioned factors, was developed: K = k{sub 1}e {sup -}ΣvQ{sub 1}/R(T+ΔT) + k{sub 2} 1/1 - α + β Φ{sup n} where K{sub 1}, K{sub 2} are the coefficients depending on the water chemistry conditions and composition of Zr alloys; α is the value of steam content; Φ is a neutron flux; n is the coefficient depending on the fuel assembly type; β is the coefficient considering the impact of impurities suppressing the radiolysis, Q{sub 1} is energy contributions of alloying components and water impurities to oxide formation, v{sub i} - stehiometry coefficient. This model allows to predict a fuel cladding corrosion taking into account the alloys composition, water chemistry and fuel burn-up. The model was verified with the help of autoclave and reactor tests for commercial and modified Zr alloys. The activation energy of oxidation process is calculating on the base of ideal mixed oxide formation model. The success of such approach makes possible to propose a generalized model for calculating the corrosion of different Zr alloys in all types of water chemistry environments of old and new LWRs. (author)

  17. Effect of Electric Field on Conductivity and Vickers Hardness of an A1-Li Alloy

    Institute of Scientific and Technical Information of China (English)

    刘兵; 陈大融; 陈铮; 王永欣; 李晓玲

    2003-01-01

    Static electric fields were applied on an aluminium-lithium alloy during solution treatment.The conductivity and Vickers hardness of the quenched Al-Li alloy is changed with the effect of electric field.The Vickers hardness increases with the applied electric field for a certain solutionizing time but decreases with the time under an electric field.In the absence of the electric field,the Vickers hardness and the conductivity increase synchronously,while reversed after electric field treatment.Positive and negative electric fields had the similar effect.The change of the local electron density in alloy caused by electric field is presented to explain the effect.

  18. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  19. Electrochemical behaviour of laser-clad Ti6Al4V with CP Ti in 0.1 M oxalic acid solution

    International Nuclear Information System (INIS)

    The relationship between the microstructure and corrosion behaviour of Ti6Al4V alloy and laser-clad commercially pure (CP) Ti coating was investigated. The microstructure, phases and properties of the clad layers were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS). Electrochemical measurement techniques including open circuit potential (OCP) and potentiodynamic polarisation were used to evaluate the corrosion behaviour of Ti6Al4V alloy in 0.1 M oxalic acid solution and the results compared to the behaviour of laser-clad CP Ti at varying laser scan speed. Results showed that laser-clad CP Ti at scan speed of 0.4 m/min formed a good cladding layer without defects such as cracks and pores. The phase present in the cladding layer was mostly α′-Ti. The microstructures of the clad layer were needle like acicular/widmanstätten α. An improvement in the microhardness values was also recorded. Although the corrosion potentials of the laser-clad samples were less noble than Ti6Al4V alloy, the polarisation measurement showed that the anodic current density was lower and also increases with increasing laser scanning speed. - Highlights: • The microstructure and corrosion behaviour of laser-clad CP Ti was investigated. • Laser-clad CP Ti 0.4 m/min scan speed gave a good coating without cracks and pores. • The phase present in the clad layer was mostly α′-Ti. • An improvement in the microhardness values was also recorded. • Anodic current density for coatings increases with increasing laser scan speed

  20. Electrochemical behaviour of laser-clad Ti6Al4V with CP Ti in 0.1 M oxalic acid solution

    Energy Technology Data Exchange (ETDEWEB)

    Obadele, Babatunde Abiodun, E-mail: obadele4@gmail.com [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Olubambi, Peter A. [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Andrews, Anthony [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Department of Materials Engineering, Kwame Nkrumah University of Science and Technology, Kumasi (Ghana); Pityana, Sisa [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); National Laser Center, Council for Scientific and Industrial Research, Pretoria (South Africa); Mathew, Mathew T. [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Department of Orthopedic Surgery, Rush University Medical Center, Chicago, IL 60612 (United States)

    2015-10-15

    The relationship between the microstructure and corrosion behaviour of Ti6Al4V alloy and laser-clad commercially pure (CP) Ti coating was investigated. The microstructure, phases and properties of the clad layers were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS). Electrochemical measurement techniques including open circuit potential (OCP) and potentiodynamic polarisation were used to evaluate the corrosion behaviour of Ti6Al4V alloy in 0.1 M oxalic acid solution and the results compared to the behaviour of laser-clad CP Ti at varying laser scan speed. Results showed that laser-clad CP Ti at scan speed of 0.4 m/min formed a good cladding layer without defects such as cracks and pores. The phase present in the cladding layer was mostly α′-Ti. The microstructures of the clad layer were needle like acicular/widmanstätten α. An improvement in the microhardness values was also recorded. Although the corrosion potentials of the laser-clad samples were less noble than Ti6Al4V alloy, the polarisation measurement showed that the anodic current density was lower and also increases with increasing laser scanning speed. - Highlights: • The microstructure and corrosion behaviour of laser-clad CP Ti was investigated. • Laser-clad CP Ti 0.4 m/min scan speed gave a good coating without cracks and pores. • The phase present in the clad layer was mostly α′-Ti. • An improvement in the microhardness values was also recorded. • Anodic current density for coatings increases with increasing laser scan speed.

  1. Effects of process variables on the burst properties of the PHWR fuel clad tubes

    International Nuclear Information System (INIS)

    Zirconium alloy tubing is used to clad the natural uranium oxide fuel in nuclear reactors. The reliability of zircaloy fuel pin depends largely on the durability of cladding under pressure of fission gasses, thermal gradients and effects of neutron bombardments (embrittlements and swelling from irradiation). To ensure the largest possible service, it is necessary to scrupulously inspect the tubes to eliminate the manufacturing defects, which might cause their premature failure in nuclear reactors. Hence, metallurgical, chemical, and mechanical properties are evaluated carefully during hot working and cold working stages. The fuel cladding which is not subjected solely to axial stress, but to bi-axial stresses imposed on the cladding by pressurized coolant, by the thermal expansion of uranium dioxide fuel and at high burn up by fuels swelling. For stresses other than those produced by the pressurized coolant, the actual longitudinal to tangential stress ratio is a variable, depending on the fuel design. Since the stress ratio can have a significant effect on the mechanical properties and because of the anisotropic nature of zirconium alloys, proper assessments of the mechanical properties for fuel cladding can best be accomplished by a test which imposes a bi-axial stress on the tubing. Many ways of testing have been tried to assess the transverse properties. There are two main groups, burst tests and ring tests. The burst tests are used widely, because of the well defined testing conditions, while ring tests although simple are not so well accepted, because of inherently ambiguous conditions for plastic instability. (author)

  2. High temperature oxidation experiments with sponge base E110G cladding

    International Nuclear Information System (INIS)

    High temperature oxidation experiments with sponge base E110G alloy were performed in wide range of parameters to investigate the oxidation behaviour of this fuel cladding in steam and in hydrogen rich steam environment; furthermore to study the susceptibility of this alloy to breakaway phenomenon. These tests are part of a systematic investigation of E110G cladding in order to facilitate the licensing of new cladding for Paks Nuclear Power Plant, in Hungary. The oxidation tests were carried out in the temperature range of 600–1200 °C under isothermal conditions. The new and the traditional types of cladding ring were compared. The experimental results showed similar behaviour of E110G and E110 samples in most of the temperature. However, the oxidation of E110 was significantly faster at 900 and at 1000 °C due to the breakaway oxidation. The oxide layer of the E110 cladding became spalling in contrast to the intact oxide layer of the new E110G cladding. The hydrogen content of the oxidised claddings was measured. Only a very small amount of hydrogen (below 100 wppm) was detected in samples of E110G, because the absorption of hydrogen was limited by the compact oxide layer. The presence of breakaway oxidation was investigated in steam atmosphere by on-line hydrogen detection between 800 and 1200 °C. No breakaway oxidation of E110G was observed during the tests up to 2700 s. Test series was carried out in steam-hydrogen mixture in the temperature range of 900–1100 °C. Hydrogen rich environment had no significant effect on the E110G oxidation. (authors)

  3. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  4. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  5. Screening of advanced cladding materials and UN-U3Si5 fuel

    Science.gov (United States)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  6. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  7. Physico Chemistry of the Chlorination of Aluminum Claddings in the Framework of HALOX Project

    International Nuclear Information System (INIS)

    The conditioning of spent nuclear fuels from test and research reactors requires a previous physicochemical treatment to stabilize them chemically.A possible way of processing is through what was called in CNEA as Process HALOX (Halogenation and Oxidation).It consists of the selective separation of cladding by halogenation and the subsequent oxidation of the core, previously to insert it into a vitreous matrix.The halogenation aim is to transform the constituents of the 6061aluminum alloy into volatile halides.In this work we present preliminary results of the chlorination of two aluminum alloys: AA 6061 and a type of CuZnAl alloy

  8. Three-Point Bending Test for Metal-Ceramic Hybrid Fuel Cladding Tubes

    International Nuclear Information System (INIS)

    A metal-ceramic hybrid cladding tube is one of these concepts. The tube forms composite ceramic layers on the zirconium alloy cladding tubes to enhance their stability during all accidents, as well as to prevent them from generating hydrogen gas under severe accidents. The new tubes can be made in the following two stages: first, producing zirconium alloy tubes, and second, forming ceramic composites on the tubes. The first stage of the fabrication is the same as a conventional manufacturing process for zirconium fuel cladding tubes. The inner tube was produced from a Zr-alloy ingot by repeating the pilgering and annealing. The second stage covers the fabrication of a SiC ceramic fiber composite. The SiC-fiber filament is wound on the Zr tube. The hetero interface between Zr metal and SiC ceramic can be adjusted with compliant media to minimize the material incoherence. The empty space within the fiber-wound perform is then filled with SiC-based preceramic polymer. Finally, the cladding tube is completed with surface coating

  9. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1997-12-01

    The author has established a range of compositions for these alloys within which hot cracking resistance is very good, and within which cold cracking can be avoided in many instances by careful control of welding conditions, particularly preheat and postweld heat treatment. For example, crack-free butt welds have been produced for the first time in 12-mm thick wrought Fe{sub 3}Al plate. Cold cracking, however, still remains an issue in many cases. The author has developed a commercial source for composite weld filler metals spanning a wide range of achievable aluminum levels, and are pursuing the application of these filler metals in a variety of industrial environments. Welding techniques have been developed for both the gas tungsten arc and gas metal arc processes, and preliminary work has been done to utilize the wire arc process for coating of boiler tubes. Clad specimens have been prepared for environmental testing in-house, and a number of components have been modified and placed in service in operating kraft recovery boilers. In collaboration with a commercial producer of spiral weld overlay tubing, the author is attempting to utilize the new filler metals for this novel application.

  10. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    Energy Technology Data Exchange (ETDEWEB)

    R. Schreiner

    2004-10-21

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database.

  11. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  12. X65/2205耐蚀合金内衬管焊接工艺开发及焊接接头耐蚀性研究%Welding Procedure Development and Corrosion Resistance Research of Welded Joints for X65/2205 Anticorrosion Alloy Clad Pipe

    Institute of Scientific and Technical Information of China (English)

    张念涛; 徐连勇; 韩永典; 刘永贞; 许可望

    2012-01-01

    根据X65/2205耐蚀合金内衬管焊接特点,开发了一种钨极氩弧焊焊接工艺.对内衬管焊接接头进行了显微组织分析、点蚀试验,抗H2S应力腐蚀开裂试验(SSCC)分析.结果表明:焊接接头HAZ的显微组织靠母材一侧为粗大的奥氏体晶粒,靠焊缝一侧为奥氏体基体上分布着铁素体;根部焊缝的显微组织为奥氏体基体上分布着铁素体.经过72h的点蚀试验后,对根部焊缝而言,钨极氩孤焊背部免充气保护焊接工艺得到的根部焊缝的耐腐蚀性与相邻的根部母材相当.经过720 h SSCC试验后,焊接接头均未发生开裂.%According to the welding characteristics of X65/2205 anticorrosion alloy clad pipe, a GTAW welding technology was developed. The microstructure analysis, pitting corrosion tests and sulfide stress corrosion crack (SSCC) tests were conducted on the X65/2205 clad pipe welded joints. The results revealed that the microstructure of HAZ was coarse austenite near base metal side and ferrite distributed in the austenite matrix near the weld metal side. The microstructure in the weld metal was ferrite distributed in the austenite matrix. After 72 h pitting corrosion test , the corrosion resistance of root weld metal, which was obtained by adopting TIG welding technology of non-filling argon in the back, the corrosion resistance was commensurate to that of the adjacent base metal. After 720 h SSCC test, the result showed that no crack was observed in the welded joints.

  13. Bioactivity of calcium phosphate bioceramic coating fabricated by laser cladding

    Science.gov (United States)

    Zhu, Yizhi; Liu, Qibin; Xu, Peng; Li, Long; Jiang, Haibing; Bai, Yang

    2016-05-01

    There were always strong expectations for suitable biomaterials used for bone regeneration. In this study, to improve the biocompatiblity of titanium alloy, calcium phosphate bioceramic coating was obtained by laser cladding technology. The microstructure, phases, bioactivity, cell differentiation, morphology and resorption lacunae were investigated by optical microscope (OM), x-ray diffraction (XRD), methyl thiazolyl tetrazolium (MTT) assay, tartrate-resistant acid phosphatase (TRAP) staining and scanning electronic microscope (SEM), respectively. The results show that bioceramic coating consists of three layers, which are a substrate, an alloyed layer and a ceramic layer. Bioactive phases of β-tricalcium phosphate (β-TCP) and hydroxyapatite (HA) were found in ceramic coating. Osteoclast precursors have excellent proliferation on the bioceramic surface. The bioceramics coating could be digested by osteoclasts, which led to the resorption lacunae formed on its surface. It revealed that the gradient bioceramic coating has an excellent bioactivity.

  14. Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

    International Nuclear Information System (INIS)

    Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

  15. Flat-Cladding Fiber Bragg Grating Sensors for Large Strain Amplitude Fatigue Tests

    OpenAIRE

    Xijia Gu; Cheng Li; Aihen Feng; Daolun Chen

    2010-01-01

    We have successfully developed a flat-cladding fiber Bragg grating sensor for large cyclic strain amplitude tests of up to ±8,000 με. The increased contact area between the flat-cladding fiber and substrate, together with the application of a new bonding process, has significantly increased the bonding strength. In the push-pull fatigue tests of an aluminum alloy, the plastic strain amplitudes measured by three optical fiber sensors differ only by 0.43% at a cyclic strain amplitude of ±7,000 ...

  16. Corrosion tests of candidate fuel cladding and reactor internal structural materials

    International Nuclear Information System (INIS)

    Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor (SCWR) in static and flowing supercritical water (SCW) autoclave at the temperatures of 550, 600 and 650°C, pressure of about 25MPa, deaerated or saturated dissolved hydrogen (STP). Samples are nickel base alloy type Hastelloy C276, austenitic stainless steels type 304NG and AL-6XN, ferritic/martensitic (F/M) steel type P92, and oxide dispersion strengthened steel MA 956. This paper focuses on the formation and breakdown of corrosion oxide scales, and proposes the future trend for the development of SCWR fuel cladding materials. (author)

  17. Corrosion tests of candidate fuel cladding and reactor internal structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, L.; Zhu, F.; Bao, Y. [Shanghai Jiao Tong Univ., School of Nuclear Science and Engineering, Shanghai (China); Tang, R. [Nuclear Power Inst. of China, National Key Lab. for Nuclear Fuel and Materials, Chengdu, Sichuan (China)

    2010-07-01

    Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor (SCWR) in static and flowing supercritical water (SCW) autoclave at the temperatures of 550, 600 and 650°C, pressure of about 25MPa, deaerated or saturated dissolved hydrogen (STP). Samples are nickel base alloy type Hastelloy C276, austenitic stainless steels type 304NG and AL-6XN, ferritic/martensitic (F/M) steel type P92, and oxide dispersion strengthened steel MA 956. This paper focuses on the formation and breakdown of corrosion oxide scales, and proposes the future trend for the development of SCWR fuel cladding materials. (author)

  18. Frictional Behavior of Fe-based Cladding Candidates for PWR

    International Nuclear Information System (INIS)

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  19. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  20. Microstructure and hardness of Mg–9Li–6Al–xLa (x=0, 2, 5) alloys during solid solution treatment

    International Nuclear Information System (INIS)

    The microstructure evolution of Mg–9Li–6Al–xLa (x=0, 2, 5) alloy under different solid solution parameters was investigated. The results show that, during solution treatment at 350 °C, the lamellar AlLi is precipitated from α-Mg in Mg–9Li–6Al, while the MgLi2Al is dissolved into the matrix. However, during solution treatment at 450 °C, the AlLi phase is wholly dissolved into matrix, while the MgLi2Al is precipitated from β-Li. The addition of La can reduce the size of α-Mg, restrain the formation of AlLi, and make the precipitated MgLi2Al from β-Li at 450 °C be finer than that in Mg–9Li–6Al. With the addition of La, the decrease of the amount of AlLi and MgLi2Al leads to a descent of hardness, while the refinement, Al–La phase precipitation, and the solution of Al atoms can improve the hardness of the alloys

  1. Eddy-Current Testing of Finned Fuel Cladding

    International Nuclear Information System (INIS)

    Eddy-current methods of testing reactor-fuel components are well established. The literature, however, mainly describes tests which are applied to simple geometries such as cylindrical rods or tubes. Recent AECL fuel designs have called for cladding with heat transfer or locating fins along the length of the fuel. This paper describes the application of eddy-current techniques to three such designs. The function and geometry of the fins must be considered in the selection of the optimum test parameters and the most suitable test coil geometry. Thus, the presence of fins may limit or restrict the test but they will not prevent a successful test. Where the fin geometry is complex eddy currents may well be the most suitable of the non-destructive methods which can be used for flaw detection. The thickness of aluminium cladding over a uranium core is measured with a small probe coil placed between the fins and shielded from them. Two flaw detection tests are described, one on sintered aluminium product (SAP) tubing using an internal bobbin coil and the other on an aluminium-clad uranium-aluminium alloy rod with an external encircling coil. The instrumentation described is relatively simple. A small portable instrument was designed for the cladding thickness measurement. For flaw detection a standard oscilloscope with a plug-in carrier-amplifier module provides a means of sensing and displaying the test coil impedance variations. This equipment ,although it does not permit sophisticated methods of eliminating unwanted noise is adequate for a variety of testing applications and has been specified for routine fuel testing on a production basis. (author)

  2. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  3. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO{sub 2}–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  4. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  5. Dilution and Ferrite Number Prediction in Pulsed Current Cladding of Super-Duplex Stainless Steel Using RSM

    Science.gov (United States)

    Eghlimi, Abbas; Shamanian, Morteza; Raeissi, Keyvan

    2013-12-01

    Super-duplex stainless steels have an excellent combination of mechanical properties and corrosion resistance at relatively low temperatures and can be used as a coating to improve the corrosion and wear resistance of low carbon and low alloy steels. Such coatings can be produced using weld cladding. In this study, pulsed current gas tungsten arc cladding process was utilized to deposit super-duplex stainless steel on high strength low alloy steel substrates. In such claddings, it is essential to understand how the dilution affects the composition and ferrite number of super-duplex stainless steel layer in order to be able to estimate its corrosion resistance and mechanical properties. In the current study, the effect of pulsed current gas tungsten arc cladding process parameters on the dilution and ferrite number of super-duplex stainless steel clad layer was investigated by applying response surface methodology. The validity of the proposed models was investigated by using quadratic regression models and analysis of variance. The results showed an inverse relationship between dilution and ferrite number. They also showed that increasing the heat input decreases the ferrite number. The proposed mathematical models are useful for predicting and controlling the ferrite number within an acceptable range for super-duplex stainless steel cladding.

  6. Elastic-plastic deformation of a nuclear fuel cladding specimen under the internal pressure of a polymer pellet

    International Nuclear Information System (INIS)

    Full text: During the operation of light water reactors, corrosion results into the development of an oxide layer, on the external surface of zirconium alloy fuel cladding, and the introduction of hydrogen into the metal (Zr+2H2O→ZrO2+2H2). Initially, hydrogen is in solid solution and diffuses towards regions of low hydrogen concentration, high hydrostatic stress and low temperature. However, with increasing time of reactor operation, the hydrogen concentration may exceed its terminal solid solubility and brittle zirconium hydrides may precipitate. Indeed hydrides are present in high burn-up fuel cladding, which is therefore more susceptible to failure, depending on hydride volume fraction and existing defects in the oxide layer. The expansion due to compression (EDC) test has been developed for the study of irradiated and hydrided cladding failure, under high hoop strain rates, which are expected during a reactivity initiated accident (RIA). During this test, a piece of cladding tube is circumferentially loaded in tension due to the expansion of a polymer pellet, axially compressed inside the tube. A finite element simulation of the EDC-test is discussed. The objective of the study is: (i) to understand the deformation of the cladding, during the experiment, including the effect of cladding material properties, and (ii) to provide information, necessary for the development of failure criteria. The distributions of important field quantities with respect to the damage of the cladding are derived together with the evolution of their maximum values, during loading. It is shown that, before cladding yielding as well as after substantial plastic deformation, the radial displacement, on the external surface, and the total energy per unit volume, when appropriately normalized, vary along the cladding axis according to specific distributions, which do not depend on the level of loading. This characteristic of cladding deformation is useful for the interpretation of

  7. Effect of rare earth oxide on the properties of laser cladding layer and machining vibration suppressing in side milling

    International Nuclear Information System (INIS)

    Highlights: • A novel laser cladding powder is developed which can reduce the machining vibration. • The machining vibrations of coating are reduced and the chatter is avoided occurring. • The vibration-suppressing mechanism is analyzed. • The hardness and wear resistance of coatings are improved significantly. - Abstract: Laser cladding, which can increase the hardness and wear resistance of the used components, is widely used in remanufacture and sustainable manufacturing field. Generally, laser cladding layer should to be machined to meet the function as well as the assembly requirements. Milling is an effective mean for precision machining. However, there exist great differences of physical and mechanical performances between laser cladding layer and substrate material, including microstructure, hardness, wear resistance, etc. This produces some new milling problems for laser cladding layer, such as machining vibration which may lead to low productivity and worse surface integrity. Thus, it is necessary to develop a novel laser cladding powder which can improve the surface hardness and wear resistance, while reducing the machining vibration in milling. Laser cladding layer was prepared by FeCr alloy and La2O3 mixed powder. The effect of La2O3 on the coating properties was investigated. Signal analysis methods of the time and frequency domain were used to evaluate the effect of the La2O3 on machining vibration in the side milling laser cladding layer. The key findings of this study are: (a) with the La2O3 content increasing, the grain size decreases dramatically and the microstructure of laser cladding layer are refine; (b) the hardness and wear resistance of the coatings with La2O3 are improved significantly; and (c) the machining vibrations of laser cladding layer with La2O3 are obviously reduced and the chatter is effectively avoided occurring

  8. Optimization of Ni-Based WC/Co/Cr Composite Coatings Produced by Multilayer Laser Cladding

    Directory of Open Access Journals (Sweden)

    Andrea Angelastro

    2013-01-01

    Full Text Available As a surface coating technique, laser cladding (LC has been developed for improving wear, corrosion, and fatigue properties of mechanical components. The main advantage of this process is the capability of introducing hard particles such as SiC, TiC, and WC as reinforcements in the metallic matrix such as Ni-based alloy, Co-based alloy, and Fe-based alloy to form ceramic-metal composite coatings, which have very high hardness and good wear resistance. In this paper, Ni-based alloy (Colmonoy 227-F and Tungsten Carbides/Cobalt/Chromium (WC/Co/Cr composite coatings were fabricated by the multilayer laser cladding technique (MLC. An optimization procedure was implemented to obtain the combination of process parameters that minimizes the porosity and produces good adhesion to a stainless steel substrate. The optimization procedure was worked out with a mathematical model that was supported by an experimental analysis, which studied the shape of the clad track generated by melting coaxially fed powders with a laser. Microstructural and microhardness analysis completed the set of test performed on the coatings.

  9. Development of vanadium fuel cladding for Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Vanadium alloys are promising material for some core components of the Sodium Fast Reactors, especially for fuel cladding applications. With good mechanical properties up to 800°C at least, good behavior under irradiation above 400°C and limited swelling, they also have the benefit from fusion program. In 2010, CEA launched the manufacturing of a V-4Cr-4Ti alloy, well documented in literature, to validate the uneasy fabrication process linked to interstitial element sensitivity and potential pollution in master alloys. 30kg of CEA-J57 alloy (7 mm-plates) were fabricated for the CEA by GfE Metalle und Materialien GmbH, Nuremberg, Germany. The program includes the investigation of recrystallization, resulting microstructure and DBTT values, high temperature mechanical properties such as tensile strength and creep resistance, chemical compatibility with both the oxide fuel and the coolant and assessment of tube fabrication, actually a triplex tube with inner and outer liners to protect vanadium from oxidation during the hot processing. (author)

  10. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    International Nuclear Information System (INIS)

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al3Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a ceramic

  11. 激光熔覆原位生成NbC/Ni45合金涂层组织与性能的研究%Investigation of Microstructure and Poperties of NbC/Ni45 Alloy Composite Coating by Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    谢颂京; 董刚

    2012-01-01

    The in situ synthesized NbC particles reinforced Ni-based alloy composite coating has been successfully prepared on 1045 steel substrate by laser cladding. The coating is free of pores and cracks with excellent bonding between the coating and the substrate. The microstructure of the coating is mainly composed of γ-Ni dendrite, a large amount of interdendritic eutectics, M23 (CB)6 type carbides and dispersed NbC particles. The growth mechanism of the NbC particles with cores is nucleation-growth and the un-melted niobium may act as the nucleation core for NbC. Compared to the pure Ni-based alloy coating, the hardness of the composite coating is increased about 36 %, giving a high average hardness of approximate HV0.2750. This is attributed to the presence of in situ synthesized NbC particles and their well distribution in the coating.%利用C02激光器在45#钢基体上成功制备了原位生成NbC颗粒增强的镍基合金涂层,涂层与基体呈现良好的冶金结合,无裂纹气孔等缺陷.涂层组织主要有γ-Ni树枝晶,枝晶间大量的共晶组织,M23 (CB)6型碳化物和弥散分布的原位生成的NbC颗粒组成.带核的NbC颗粒是以为完全溶解的Nb为核心在其上长大的.由于原位生成NbC颗粒在复合涂层中的均匀分布,使涂层的平均显微硬度高达HV0.2750,比纯Ni45合金涂层提高了约36%.

  12. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes the manufacturing technology developed for HANATM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANATM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANATM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANATM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANATM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANATM Lead Test Rods(LTR) in a commercial reactor

  13. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    International Nuclear Information System (INIS)

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  14. Investigation of research and development subjects for very high burnup fuel. Development of fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Nagase, Fumihisa; Suzuki, Masahide; Furuta, Teruo; Suzuki, Yasufumi; Hayashi, Kimio; Amano, Hidetoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-05-01

    Plutonium use as well as burnup extension of UO{sub 2} fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a `very high burnup` aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs.

  15. Comparison of fuel performance codes using stainless steel as cladding material

    International Nuclear Information System (INIS)

    The cladding material in the firsts PWRs (pressurized water reactors) was stainless steel which was later replaced by zirconium-based alloys mainly due to their lower neutron absorption cross section which impacts on the operational cost of the reactor. However, after Fukushima Daiichi Accident, stainless steel cladding appears as an alternative to zirconium-based alloys to overcome the safety problems related to hydrogen production and explosion in nuclear power plants under severe accident conditions. In order to assess the fuel performance of fuel rods manufactured using stainless steel as cladding material, it is necessary to modify conventional fuel performance codes developed in the last decades. The aim of this paper is to discuss the results obtained under steady-state irradiation for fuel rods manufactured using stainless steel as cladding applying two new versions of fuel performance codes, one based on FRAPCON (named IPEN-CNEN/SS) and TRANSURANUS. These fuel performance codes were modified with the inclusion of the mechanical and physical properties of the stainless steel 348. The performance under the same power history was investigated and compared considering the stainless steel expected behavior. The results obtained from IPEN-CNEN/SS and TRANSURANUS simulations have presented the same trends and global performance under steady-state irradiation. (author)

  16. Technical development of double-clad process for thin strip casting of carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, H.L.; Forkel, C.E.; Knudson, D.L.

    1984-08-01

    This report documents the technical development for a patent disclosure of a double-clad process for the continuous casting of thin-strip carbon steel. The fundamental idea of the disclosure is to form a product strip by depositing molten steel between two, cooled, clad strips of the same material. The claimed benefits include: (a) the conservation of energy in steel making through the elimination of soaking pits and reheat cycles, and (b) an improved surface on both sides of the as-cast product such that it will be suitable for direct feed to a cold-reduction mill. However, the process as conceived is not necessarily limited to the casting of carbon steel, but may be also applied to other metals and alloys. The work is described under three headings as follows. Preliminary Considerations and Scoping Analysis presents the basic idea of the double-clad, thin-strip casting process; the energy conservation potential; scoping heat transfer calculations for the casting process; and independent review of this work. Thermal Analysis for Roller Configuration of Double-Clad Process, presents the development, results, and independent review of a finite-element thermal analysis for the casting process as originally conceived (using only chilled rollers in direct contact with the clad material of the product strip). Further Considerations for Belt Configuration of Double-Clad Process deals with a modified equipment design which interposes two product support belts, one on each side of the product, between the clad strip and the rollers. In addition to the process description, this section presents the preliminary mechanical calculations for the endless metal belts and the work scope and results for the computer model revision and thermal analysis for the modified concept.

  17. Hydrogen Effect on the Circumferential Mechanical Properties of HANA-4 and HANA-6 Cladding Tubes

    International Nuclear Information System (INIS)

    KAERI has been doing a lot of out-of pile tests including an in-pile test to verify the performance of HANA cladding tubes for a high burn-up fuel rod, developed by them. When a zirconium alloy is used in a nuclear reactor, hydrides form in it from not only external hydrogen sources such as a waterside corrosion, dissolved hydrogen in a coolant, water radiolysis but also internal sources such as the hydrogen content in fuel pellets and the moisture absorbed by a uranium dioxide fuel pellet. Hydrides may act as a sudden failure at a very low strain. For low and medium hydrogen content, the hydrides crack during a tensile loading and accelerate the ductile fracture process. As a kind of simulation test to obtain the estimated data of HANA cladding tubes in a high burn-up state, the hydrogen effect on the axial tensile properties of a HANA-4(Zr-1.5Nb-0.0.4Sn- 0.21Fe-0.1Cu) cladding tube and that on the burst properties of HANA-4 and HANA-6 (Zr-1.1Nb-0.05Cu) cladding tubes was already studied. This study was also done to characterize the effect of hydrogen on the circumferential mechanical properties of HANA-4 and HANA-6 cladding tubes by a ring tension test at both room temperature and 350 .deg. C. Additional tests were also done on both Zircaloy-4 (Zr-1.26Sn-0.23Fe-0.12Cr) and A (Zr-1.0Nb-0.99Sn-0.11Fe) cladding tubes of a commercial grade to compare the hydrogen effect on their circumferential properties with that on the properties of the HANA-4 and HANA-6 cladding tubes

  18. Constraint effects of clad on underclad crack

    International Nuclear Information System (INIS)

    The finite element method is applied to two-dimensional elastic-plastic analyses for underclad crack problems. The analyses are performed for rectangular specimens with an underclad crack, which are composed of A533B class 1 steel and a clad material, to obtain the fracture mechanics parameter J-integral and the stress distribution ahead of a crack tip. The Q-factor proposed by O'Dowd and Shih is calculated from the stress distribution ahead of a crack tip, and the constraint effect of a crack tip due to a clad material or the effect of a clad material on the fracture toughness of a base material is discussed in terms of the Q-factor. Clad thickness, crack length and the material property of a clad material are varied to examine their effects

  19. Abrasive Performance of Chromium Carbide Reinforced Ni3Al Matrix Composite Cladding

    Institute of Scientific and Technical Information of China (English)

    LI Shang-ping; LUO He-li; FENG Di; CAO Xu; ZHANG Xi-e

    2009-01-01

    The Microstructure and room temperature abrasive wear resistance of chromium carbide reinforced NiM3Al matrix composite cladding at different depth on nickel base alloy were investigated. The results showed that there is a great difference in microstructure and wear resistance of the Ni3 Al matrix composite at different depth. Three kinds of tests, designed for different load and abrasive size, were used to understand the wear behaviour of this material. Under all three wear conditions, the abrasion resistance of the composite cladding at the depth of 6 mm, namely NC-M2, was much higher than that of the composite cladding at the depth of 2 mm, namely NC-M1. In addition, the wear-resistant advantage of NC-M2 was more obvious when the size of the abrasive was small. The relative wear resistance of NC-M2 increased from 1.63 times to 2.05 times when the size of the abrasive decreased from 180 μm to 50μm. The mierostructure of the composite cladding showed that the size of chromium carbide particles, which was mainly influenced by cooling rate of melting pool, was a function of distance from the interface between the coating and substrate varied gradually. The chromium carbide particles near the interface were finer than that far from inter-face, which was the main reason for the different wear resistance of the composite cladding at different depth.

  20. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

  1. ORNL Analysis of Operational and Safety Performance for Candidate Accident Tolerant Fuel and Cladding Concepts

    International Nuclear Information System (INIS)

    Enhanced accident-tolerant fuels (ATFs) are being developed by the US Department of Energy Office of Nuclear Energy Fuel Cycle Research and Development Program to replace standard Zircaloy cladding and/or UO2 fuel in light water reactors. Proposed ATF concepts seek to reduce severe accident (SA) risks by increasing the coping time available to operators for accident response, reducing the extent and rate of heat and hydrogen production from steam oxidation, or enhancing fission product retention. Candidate ATF concepts require analyses to demonstrate adequate performance during normal operation and worthwhile improvements in SA scenarios. Two key ATF areas are being developed at Oak Ridge National Laboratory: (1) alternate cladding materials, including advanced iron-chromium-aluminium (FeCrAl) alloys and silicon carbide (SiC) composites, and (2) fully ceramic microencapsulated (FCM) fuel, which uses coated fuel particles embedded in an SiC matrix. Reactor physics analyses examining candidate ATF clad materials in a pressurized water reactor (PWR), with preliminary assessments of combinations of fuel enrichment and cladding thickness required to match existing cycle lengths and economic factors such as fuel costs, are presented. SA analyses including updated analyses of how FeCrAl cladding and channel box impact SA scenarios in a boiling water reactor (BWR) are also discussed. (author)

  2. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Science.gov (United States)

    Courty, Olivier; Motta, Arthur T.; Hales, Jason D.

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick's law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  3. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jason D. Hales; Various

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  4. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Courty, Olivier, E-mail: o.courty@gmail.com [Pennsylvania State University, 45 Bd Gouvion Saint Cyr, 75017 Paris (France); Motta, Arthur T., E-mail: atm2@psu.edu [Department of Mechanical and Nuclear Engineering, 227 Reber Building, Penn State University, University Park, PA 16802 (United States); Hales, Jason D., E-mail: jason.hales@inl.gov [Fuels Modeling and Simulation Department, Idaho National Laboratory (United States)

    2014-09-15

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  5. AP1000 steam generator tube sheet forging cladding quality issue analysis and precaution method

    International Nuclear Information System (INIS)

    The dimension of AP1000 Steam Generator is 4500 mm × H1170 mm. Many cracks which length between 9∼ 50 mm, depth between 10∼14 mm are found through Ultrasonic Inspection after cladding. The material of AP1000 Steam Generator is ASME SA508 Grade 3 Class 2. The metallurgical structure of ASME SA508 Grade 3 Class 2 is sorbite (cubic body-centered). The low hydrogen dissolvability in cubic body-centered structure is easy resulting cracks. The cladding material is 690 nickel-base alloy which structure is face-centered cubic. Hydrogen has a bigger dissolvability in face-centered cubic structure. So if the hydrogen is not removed enough during welding, many cracks would generate. The reason why cracks generated after cladding is the preheat method and post heat temperature, post heat occasion and the high hydrogen content of forging itself. In order to obtain the qualified product after cladding, the welding process shall be decided through the theory and process analysis based on the consideration for the high hydrogen content and residual stress of the forging. The situation of cracks generated after cladding is described. The methods for precaution such cracks are provided. (author)

  6. Structure of Al-Ni alloy after equichannel angular pressing

    International Nuclear Information System (INIS)

    The structural-phase state of the Al-Li alloy rods, obtained under different conditions of the equichannel angular (ECA) pressing is studied. The fine-grained structure is formed in the ECA-pressing process, whereby the more fine grains correspond to the lower pressing temperatures. The dislocation substructure, including the subgrains, limited by the dislocation boundaries, is formed in the majority of the grains. The most developed substructure is formed in the process of pressing at the increased temperatures, when the largest grains are formed. Only the samples with such a structure manifested the superplasticity

  7. Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990/sup 0/C (1815/sup 0/F)

    Energy Technology Data Exchange (ETDEWEB)

    Slaby, J.G.; Siegel, B.L.; Gedeon, L.; Galbo, R.J.

    1973-10-01

    The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 h at 990/sup 0/C (1815/sup 0/F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100/sup 0/C (2012/sup 0/F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

  8. Underwater laser beam welding of Alloy 690

    International Nuclear Information System (INIS)

    Stress Corrosion Clacking (SCC) has been reported at Alloy 600 welds between nozzles and safe-end in Pressurized Water Reactor (PWR) plant. Alloy 690, which has higher chromium content than Alloy 600, has been applied for cladding on Alloy 600 welds for repairing damaged SCC area. Toshiba has developed Underwater Laser Beam Welding technique. This method can be conducted without draining, so that the repairing period and the radiation exposure during the repair can be dramatically decreased. In some old PWRs, high-sulfur stainless steel is used as the materials for this section. It has a high susceptibility of weld cracks. Therefore, the optimum welding condition of Alloy 690 on the high-sulfur stainless steel was investigated with our Underwater Laser Beam Welding unit. Good cladding layer, without any crack, porosity or lack of fusion, could be obtained. (author)

  9. Flat-cladding fiber Bragg grating sensors for large strain amplitude fatigue tests.

    Science.gov (United States)

    Feng, Aihen; Chen, Daolun; Li, Cheng; Gu, Xijia

    2010-01-01

    We have successfully developed a flat-cladding fiber Bragg grating sensor for large cyclic strain amplitude tests of up to ± 8,000 με. The increased contact area between the flat-cladding fiber and substrate, together with the application of a new bonding process, has significantly increased the bonding strength. In the push-pull fatigue tests of an aluminum alloy, the plastic strain amplitudes measured by three optical fiber sensors differ only by 0.43% at a cyclic strain amplitude of ± 7,000 με and 1.9% at a cyclic strain amplitude of ± 8,000 με. We also applied the sensor on an extruded magnesium alloy for evaluating the peculiar asymmetric hysteresis loops. The results obtained were in good agreement with those measured from the extensometer, a further validation of the sensor. PMID:22163621

  10. Flat-Cladding Fiber Bragg Grating Sensors for Large Strain Amplitude Fatigue Tests

    Directory of Open Access Journals (Sweden)

    Xijia Gu

    2010-08-01

    Full Text Available We have successfully developed a flat-cladding fiber Bragg grating sensor for large cyclic strain amplitude tests of up to ±8,000 με. The increased contact area between the flat-cladding fiber and substrate, together with the application of a new bonding process, has significantly increased the bonding strength. In the push-pull fatigue tests of an aluminum alloy, the plastic strain amplitudes measured by three optical fiber sensors differ only by 0.43% at a cyclic strain amplitude of ±7,000 με and 1.9% at a cyclic strain amplitude of ±8,000 με. We also applied the sensor on an extruded magnesium alloy for evaluating the peculiar asymmetric hysteresis loops. The results obtained were in good agreement with those measured from the extensometer, a further validation of the sensor.

  11. Factors affecting out-of and in-reactor corrosion of Zr claddings of fuel rods

    International Nuclear Information System (INIS)

    Results are generalized that were acquired from studying corrosion behaviour of E110 and E635 cladding under research reactor and commercial WWER conditions. The role was assessed that is played by surface boiling, heat flux and neutron irradiation in the mode of corrosion damage of E110 and E635 alloys. The results of out-of-pile tests are analyzed, i.e., the influence of environment, Li concentration of water, conditions of water and its content of oxygen. (author)

  12. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2015-05-01

    Full Text Available Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  13. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  14. Microstructure characteristics of Ni/WC composite cladding coatings

    Science.gov (United States)

    Yang, Gui-rong; Huang, Chao-peng; Song, Wen-ming; Li, Jian; Lu, Jin-jun; Ma, Ying; Hao, Yuan

    2016-02-01

    A multilayer tungsten carbide particle (WCp)-reinforced Ni-based alloy coating was fabricated on a steel substrate using vacuum cladding technology. The morphology, microstructure, and formation mechanism of the coating were studied and discussed in different zones. The microstructure morphology and phase composition were investigated by scanning electron microscopy, optical microscopy, X-ray diffraction, and energy-dispersive X-ray spectroscopy. In the results, the coating presents a dense and homogeneous microstructure with few pores and is free from cracks. The whole coating shows a multilayer structure, including composite, transition, fusion, and diffusion-affected layers. Metallurgical bonding was achieved between the coating and substrate because of the formation of the fusion and diffusion-affected layers. The Ni-based alloy is mainly composed of γ-Ni solid solution with finely dispersed Cr7C3/Cr23C6, CrB, and Ni+Ni3Si. WC particles in the composite layer distribute evenly in areas among initial Ni-based alloying particles, forming a special three-dimensional reticular microstructure. The macrohardness of the coating is HRC 55, which is remarkably improved compared to that of the substrate. The microhardness increases gradually from the substrate to the composite zone, whereas the microhardness remains almost unchanged in the transition and composite zones.

  15. High Temperature and Pressure Steam-H2 Interaction with Candidate Advanced LWR Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL

    2012-08-01

    This report summarizes the work completed to evaluate cladding materials that could serve as improvements to Zircaloy in terms of accident tolerance. This testing involved oxidation resistance to steam or H{sub 2}-50% steam environments at 800-1350 C at 1-20 bar for short times. A selection of conventional alloys, SiC-based ceramics and model alloys were used to explore a wide range of materials options and provide guidance for future materials development work. Typically, the SiC-based ceramic materials, alumina-forming alloys and Fe-Cr alloys with {ge}25% Cr showed the best potential for oxidation resistance at {ge}1200 C. At 1350 C, FeCrAl alloys and SiC remained oxidation resistant in steam. Conventional austenitic steels do not have sufficient oxidation resistance with only {approx}18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application and high Cr contents raise concern about {alpha}{prime} formation. Higher pressures (up to 20.7 bar) and H{sub 2} additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed in a H{sub 2}-50%H{sub 2}O environment at 10.3 bar. As some of the results regarding low-alloyed FeCrAl and Fe-Cr alloys were unexpected, further work is needed to fundamentally understand the minimum Cr and Al alloy contents needed for protective behavior in these environments in order to assist in alloy selection and guide alloy development.

  16. The use of slow strain rate technique for studying stress corrosion cracking of an advanced silver-bearing aluminum-lithium alloy

    International Nuclear Information System (INIS)

    In the present study, stress corrosion cracking (SCC) behavior of naturally aged advanced silver-bearing Al-Li alloy in NaCl solution was investigated using slow strain rate test (SSRT) method. The SSRT’s were conducted at different strain rates and applied potentials at room temperature. The results were discussed based on percent reductions in tensile elongation in a SCC-causing environment over those in air tended to express the SCC susceptbility of the alloy under study at T3. The SCC behavior of the alloy was also discussed based on the microstructural and fractographic examinations

  17. The use of slow strain rate technique for studying stress corrosion cracking of an advanced silver-bearing aluminum-lithium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Frefer, Abdulbaset Ali; Raddad, Bashir S. [Department of Mechanical and Industrial Engineering/Tripoli University, Tripoli (Libya); Abosdell, Alajale M. [Department of Mechanical Engineering/Mergeb University, Garaboli (Libya)

    2013-12-16

    In the present study, stress corrosion cracking (SCC) behavior of naturally aged advanced silver-bearing Al-Li alloy in NaCl solution was investigated using slow strain rate test (SSRT) method. The SSRT’s were conducted at different strain rates and applied potentials at room temperature. The results were discussed based on percent reductions in tensile elongation in a SCC-causing environment over those in air tended to express the SCC susceptbility of the alloy under study at T3. The SCC behavior of the alloy was also discussed based on the microstructural and fractographic examinations.

  18. Development of eutectic free cladding materials for metallic fuel

    International Nuclear Information System (INIS)

    Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10 mass%Zr alloy were conducted at a temperature of 1027K for 2250 hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr with layer was observed in all of these materials. Zr with layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN. (author)

  19. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Was, G S; Allen, T R; Ila, D; C,; Levi,; Morgan, D; Motta, A; Wang, L; Wirth, B

    2011-06-30

    The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that have yet to be resolved or that are beyond the existing experience base of dose or burnup. The research program consists of three major thrusts: 1) high-dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature, and 3) innovative cladding concepts embodying functionally-graded barrier materials. This NERI-Consortium final report represents the collective efforts of a large number of individuals over a period of three and a half years and included 9 PIs, 4 scientists, 3 post-docs and 12 students from the seven participating institutions and 8 partners from 5 national laboratories and 3 industrial institutions (see table). University participants met semi-annually and participants and partners met annually for meetings lasting 2-3 days and designed to disseminate and discuss results, update partners, address outstanding issues and maintain focus and direction toward achieving the objectives of the program. The participants felt that this was a highly successful program to address broader issues that can only be done by the assembly of a range of talent and capabilities at a more substantial funding level than the traditional NERI or NEUP grant. As evidence of the success, this group, collectively, has published 20 articles in archival journals and made 57 presentations at international conferences on the results of this consortium.

  20. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  1. Solidification and microstructural aspects of laser-deposited Ni–Mo–Cr–Si alloy on stainless steel

    Indian Academy of Sciences (India)

    Reena Awasthi; Santosh Kumar; D Srivastava; G K Dey

    2010-12-01

    Laser cladding of stainless steel substrate was carried out using Ni–32Mo–15Cr–3Si (wt%) alloy powder. Laser cladding parameters were optimized to obtain defect-free and metallurgically bonded clad. Variation in solidification rate, cooling rate and compositional variation resulted in heterogeneous microstructure. Microstructure was found to be distinctly different in regions of clad cross-section. Majority of the region was found to consist of eutectic of Mo-rich hcp intermetallic Laves phase and NiFe fcc gamma solid solution phases. Extensive microstructural examinations of different clad regions have been carried out using microscopy and microanalysis techniques.

  2. Development of Zirconium alloys in China

    International Nuclear Information System (INIS)

    The important program of construction of nuclear reactors in China requires self-reliance on zirconium alloys for fuel cladding and assembly structural materials. This series of slides presents China's research and development program on zirconium alloys. Experimental studies have shown that N18 and N36 zirconium alloys have excellent out-of-pile properties comparing to Zr-4. Engineering studies have been performed on the N36 alloy, these studies have involved: the N36 alloy production on industrial scale, the manufacturing of N36 tubes and bar materials for fuel cladding, the resistance to corrosion and the irradiation behavior. N36 cladding tubes meet the design requirements and are superior to Zr-4 in terms of behavior in ionized water (and 70 ppm LiOH), in terms of creep resistance, of low-cycle fatigue property and of high temperature oxidation. N36 lead fuel rods and assemblies show a good integrity and appearance after the 2. cycle irradiation, they are still being tested in Qinshan NPP. More irradiated data of N36 alloys is needed for evaluation of in-pile behavior. The N36 alloy possesses a good machinability and economical efficiency

  3. Additive Manufacturing of High-Entropy Alloys by Laser Processing

    NARCIS (Netherlands)

    Ocelik, V.; Janssen, Niels; Smith, Stefan; De Hosson, J. Th M.

    2016-01-01

    This contribution concentrates on the possibilities of additive manufacturing of high-entropy clad layers by laser processing. In particular, the effects of the laser surface processing parameters on the microstructure and hardness of high-entropy alloys (HEAs) were examined. AlCoCrFeNi alloys with

  4. Morphology control of anodic ZrO2 layer for the prevention of H2 production from Zr-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y. J.; Park, J. W.; Cho, S. O. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Since the Fukushima disaster happened, studies on accident-resistant nuclear fuel has been carried out actively. There has been an attempt to protect zircaloy fuel cladding by coating SiC. Research on producing oxide layer that can block fuel cladding from water on the surface of zircaloy fuel cladding by means of anodizing to reduce the rate of oxidation of fuel cladding at Loss Of Coolant Accident (LOCA) is an significant ongoing study subject. Applying nanostructured oxide layer to the prevention of thermal deformation of oxide layer was already suggested in our research group, the reasons of which is nanoporous structure is better than nanotube structure in terms of corrosion-resistant structure because nanotube structure can be easily peeled off. In this study, methods which are able to control morphology between nanoporous and nanotube structure were conducted by changing the anodizing conditions. Hence, Using glycerol and ammonium fluoride, Zircaloy-4 was anodized by varying water contents and applied voltage. It reveals that the alloy transition from nanoporous structure to nanotube structure can be changed by varying water contents of anodizing solution and applied voltage. Anodizing conditions determining nanoporous structure were obtained. According to the mechanism already suggested, nanoporous oxide layer that can seal the fuel cladding perfectly, and increase critical heat flux (CHF) due to large surface area is easily produced. This results obtained in this paper expected to be facilitated fabrication of accident-resistant nuclear fuel cladding.

  5. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  6. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H3PO4 at temperatures above 3000C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step

  7. Fracture predictions in Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial temperatures of 360 to 4050C for irradiated spent fuel cladding (wet pool storage) are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 8500C and stresses between 5 and 500 MPa. These maps are then combined with both the known temperature history (an exponentially decaying one) of Zircaloy fuel cladding in dry storage and a life fracture rule to predict the rupture life of the cladding in dry storage. Predictions of the deformation and fracture map methodology are shown to be in good agreement with constant stress-constant temperature data

  8. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  9. Performance of the Barrier between the Metallic Fuel and the Clad Material in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Metallic fuel has been considered as one of the most probable candidates of the fuel system in the Sodium-cooled Fast Reactor (SFR) in that it has high thermal conductivity, proliferation resistance, and good compatibility between sodium. Addition of the alloying element such as chromium, molybdenum, zirconium and titanium was applied in order to increase the solidus temperature of the uranium-plutonium alloy. Among these, uranium-plutonium alloys with the addition of 10-20% zirconium have been considered in the design of the metallic fuel in SFR. However, actinide elements in metallic fuel like uranium and plutonium react with stainless steel at a temperature above 650 .deg. C to form eutectic compounds. Such eutectic reaction reduces cladding thickness so that mechanical integrity of the cladding gradually decreases as the fuel burnup proceeds. To mitigate such a circumstance, barrier layer, which prevents both fuel and clad elements from diffusing each other, has been developed. Metallic foil made of pure metal has been suggested as a barrier and its feasibility test has been carried out. The objectives in this study are to propose several kinds of the barrier material and to verify its performance under a fuel-clad interaction situation

  10. High-resolution electron microscopy for structural and analytical investigations of binary aluminum alloys; Hochaufloesende Elektronenmikroskopie zur strukturellen und analytischen Untersuchung binaerer Aluminiumlegierungen

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, G.

    1994-12-31

    All investigations of this doctoral thesis were conducted on alloys of the systems AlAg and AlLi. Therefore the introductory Chapter I gives a report on the most important properties of these alloys in due brevity and explains the physical background of the experiments conducted. Chapter II presents experiments relating to the early stages of demixing of an oversaturated AlLi alloy. The determination of the size spectrum of the precipitations using high-resolution electron microscopy furnishes information which is essential to an understanding of the demixing process. Chapter III describes investigations into the behavior of the intermetallic {delta}`-phase in the system AlLi under electron irradiation. These experiments involve use of the electron microscope not only as an imaging instrument, but also as a source of radiation. The methodologically oriented final section of the present thesis (Chapter IV) investigates ``hollow cone wide angle dark field imaging`` as an advanced chemically sensitive imaging method which supplies information on the chemical composition of the sample on a nm scale, while requiring only little equipment. (orig)

  11. X-Ray Diffraction Studies of Structures of Be, Al, LiF, Fe+3%Si, Si, SiO2, KCl under Dynamic Pressures from 2 Gpa to 20 Gpa

    OpenAIRE

    Egorov, L.; Barenboim, A.; Mokhova, V.; Dorohin, V.; Samoilov, A.

    1997-01-01

    Currently, the only direct method to study behaviour of solid crystal substance structures under dynamic compression is method to record X-rays diffraction pictures of crystal structures under shock compression. The paper presents results of X-rays diffraction measurements conceming structural parameters of shock compressed substances at pressures higher than Hugoniot elastic limit (Be. Al, LiF, Fe+3%Si), lower than Hugoniot elastic limit (Si, SiO2, LiF) and in the area of pressures of phase ...

  12. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2014-06-01

    Full Text Available Corresponding author: Alan Tomasi, Group R&D Project Manager, Permasteelisa S.p.A., viale E. Mattei 21/23 | 31029 Vittorio Veneto, Treviso, Italy. Tel.: +39 0438 505207; E-mail: a.tomasi@permasteelisagroup.com; www.permasteelisagroup.com Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  13. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  14. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO2, which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO2. The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  15. Fabrication and Lasing Property of Yb~(3+)-doped Double-Clad Fibers with Novel Inner Cladding

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The Yb3+-doped double-clad fibers with novel inner cladding have been made by using MCVD process, solution-doping method and optical machining together. The laser power and slope efficiency of the fiber lasers are higher than 1.8W and 50% respectively.

  16. A cladding pumped Ytterbium-doped fiber laser with holey inner and outer cladding

    OpenAIRE

    Furusawa, Kentaro; Malinowski, A.N.; Price, Jonathan H.V.; Monro, Tanya M.; Jayanta K. Sahu; Nilsson, Johan; Richardson, David J

    2001-01-01

    We have fabricated an ytterbium doped all-glass double-clad large mode area holey fiber. A highly efficient cladding pumped single transverse mode holey fiber laser has been demonstrated, allowing continuous-wave output powers in excess of 1W with efficiencies of more than 80%. Furthermore both Q-switched and mode-locked operation of the laser have been demonstrated.

  17. Thermal conductance at millikelvin temperatures of woven ribbon cable with phosphor-bronze clad superconducting wires

    Science.gov (United States)

    Woodcraft, Adam L.; Ventura, Guglielmo; Martelli, Valentina; Holland, Wayne S.

    2010-08-01

    Woven Nomex® ribbon cables made up with superconducting niobium-titanium wire are used at millikelvin temperatures in many large cryogenic instruments. It is important to know how much heat in transmitted down such cables. However, the conductivity of the materials used is not well known. Another problem is that the wires are normally clad with alloys which exhibit some magnetism. This is a potential problem for instruments employing superconducting detectors. A safe non-magnetic alternative to the usual materials is phosphor-bronze clad niobium-titanium wiring. However, there is little experience with such wires. We have therefore measured the conductance of a ribbon cable made up with these wires. The measured values are in good agreement with our predictions, suggesting that the values we have used to model the cable are sufficiently accurate, and could therefore be used to predict the performance of ribbon cables using other cladding materials, so long as the conductivity of the cladding is reasonably well known. As part of our analysis, we consider the likely variation in thermal conductivity values for C51000 phosphor bronze caused by legitimate variations in composition.

  18. Utilizing clad piping to improve process plant piping integrity, reliability, and operations

    International Nuclear Information System (INIS)

    During the past four years carbon steel piping clad with type 304L (UNS S30403) stainless steel has been used to solve the flow accelerated corrosion (FAC) problem in nuclear power plants with exceptional success. The product is designed to allow ''like for like'' replacement of damaged carbon steel components where the carbon steel remains the pressure boundary and type 304L (UNS S30403) stainless steel the corrosion allowance. More than 3000 feet of piping and 500 fittings in sizes from 6 to 36-in. NPS have been installed in the extraction steam and other lines of these power plants to improve reliability, eliminate inspection program, reduce O and M costs and provide operational benefits. This concept of utilizing clad piping in solving various corrosion problems in industrial and process plants by conservatively selecting a high alloy material as cladding can provide similar, significant benefits in controlling corrosion problems, minimizing maintenance cost, improving operation and reliability to control performance and risks in a highly cost effective manner. This paper will present various material combinations and applications that appear ideally suited for use of the clad piping components in process plants

  19. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  20. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature

    International Nuclear Information System (INIS)

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases (α(O) and 'ex β') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the β-->α phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials were mechanically tested at

  1. ELECTROCHEMICAL STUDIES OF MOBILE CHARGED SPECIES DURING ZIRCONIUM ALLOY OXIDATION

    OpenAIRE

    Smith, James Stephen

    2013-01-01

    This research has used a suite of electrochemical techniques, both in-situ and ex-situ to investigate the mobile charged species in the oxides of zirconium alloys. Limits on the corrosion resistance of existing zirconium alloys used for fuel cladding are a major restriction on the burn-up that can be achieved within a pressurised water reactor (PWR). Developing a full mechanistic understanding of the corrosion process of zirconium alloys in the primary water environment is necessary for prolo...

  2. Determination of Stress-Corrosion Cracking in Aluminum-Lithium Alloy ML377

    Science.gov (United States)

    Valek, Bryan C.

    1995-01-01

    The use of aluminum-lithium alloys for aerospace applications is currently being studied at NASA Langley Research Center's Metallic Materials Branch. The alloys in question will operate under stress in a corrosive environment. These conditions are ideal for the phenomena of Stress-Corrosion Cracking (SCC) to occur. The test procedure for SCC calls for alternate immersion and breaking load tests. These tests were optimized for the lab equipment and materials available in the Light Alloy lab. Al-Li alloy ML377 specimens were then subjected to alternate immersion and breaking load tests to determine residual strength and resistance to SCC. Corrosion morphology and microstructure were examined under magnification. Data shows that ML377 is highly resistant to stress-corrosion cracking.

  3. Investigation of Microstructure in Solid State Welded Al-Cu-Li alloy

    Directory of Open Access Journals (Sweden)

    No Kookil

    2016-01-01

    Full Text Available Al-Li alloys have been extensively used in aerospace vehicle structure since the presence of lithium increases the modulus and reduce the density of the alloy. Especially the third generation Al-Cu-Li alloy shows enhanced fracture toughness at cryogenic temperatures so that the alloy has been used on the fuel tank of space launchers, like Super Lightweight External Tank of the Space Shuttle. Since the commercial size of the plate cannot accommodate the large tank size of the launcher, joining several pieces is required. However, lithium is highly reactive and its compounds can decompose with heat from conventional fusion welding and form different types of gases which result in formation of defects. In this study, the microstructure change is investigated after solid state welding process to join the Al-Cu-Li sheets with optical and transmission electron microscopic analysis of precipitates.

  4. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  5. Rapid manufacturing by laser sintering and laser cladding; Rapid Manufacturing durch Lasersintern und 3D-Laserstrahl-Auftragschweissen

    Energy Technology Data Exchange (ETDEWEB)

    Haferkamp, H. [Laser Zentrum Hannover e.V. (Germany); Alvensleben, F. von [Laser Zentrum Hannover e.V. (Germany); Gerken, J. [Laser Zentrum Hannover e.V. (Germany)

    1995-06-01

    Among the technologies which are under development for the direct production of metal components, the laser-supported techniques laser sintering and laser cladding offer positive expectations for industrial use. Founded on extensive work in the field of laser cladding of functional layers [1,2], results have been gathered at the Laser Zentrum Hannover (LZH) concerning the direct manufacturing of metal parts by laser supported techniques [3,4]. The different processes and first results concerning the build-up of metal parts mainly by laser sintering are described in this paper. During the investigation, the suitability of metals such as copper, nickel, aluminium and aluminium-bronze alloy for laser sintering without binders was tested. In addition, metal parts produced by laser cladding and a possibility of process monitoring are shown. For more details see 5 Extended Abstract. (orig.)

  6. Effects of homogenization treatment on the microstructure and mechanical properties of Mg–8Li–3Al–Y alloy

    International Nuclear Information System (INIS)

    Highlights: • The evolution of microstructure and phases are observed and analyzed. • The effects of homogenization on mechanical properties are investigated. • The optimum homogenization parameter for the alloy is suggested. - Abstract: The effects of homogenization treatment of Mg–8Li–3Al–Y alloy were investigated by optical microscope (OM), scanning electron microscope (SEM), X-ray diffraction (XRD), energy dispersive spectroscopy (EDS) and universal testing machine. The experiment results indicated that there existed five phases in the as-cast alloy: α, β, Al2Y, AlLi and MgAlLi2. The spheroidized α phase grew gradually and its microstructure and composition became homogeneous when it was treated at 300 °C for 12 h. Meanwhile, the alloy has a good comprehensive mechanical property compared with other homogenization schedules. So, a homogenization treatment at 300 °C for 12 h was determined to be the optimal homogenization treatment for Mg–8Li–3Al–Y alloy

  7. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 3200C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO2, respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  8. Corrosion of research reactor aluminium-clad spent fuel at wet storage

    International Nuclear Information System (INIS)

    Full text: About 700 research reactors (RRs) and critical assemblies have been built worldwide. Of these about 260 are presently operational. A large variety of spent fuels with different fuel meats, different geometries and different enrichments in 235U are presently stored under water in basins located throughout the world while awaiting final disposition or shipment to the USA or to the Russian Federation. At some reactor sites, the spent fuel has been in water for up to 40 years. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are susceptible to corrosion in water of less than optimum quality. Some fuel is stored in reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some in away-from-reactor pools. Corrosion induced degradation of fuel cladding, the underlying fuel itself, other aluminium components of the fuel handling and storage systems, especially pool liners, has been observed in many of the pools. As a result of these corrosion issues, the IAEA implemented in 1996 a Coordinated Research Project (CRP) on 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The objectives of the CRP were to: (a) establish uniform practices for corrosion monitoring and surveillance, (b) provide a technical basis for continued wet storage of RR spent fuel, (c) collect data to help predict fuel cladding lifetimes and (d) establish procedures for characterization of water in fuel storage basins. Phase-I of this CRP ended in 2000 and Phase-II is ongoing with 8 participating countries, including Kazakhstan. The CRP consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating countries and evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage basin water. The materials selected for testing were representative of typical Al cladding alloys used in RR fuel, handling tools and storage racks. The

  9. The fatigue response of the aluminium-lithium alloy, 8090

    Science.gov (United States)

    Birt, M. J.; Beevers, C. J.

    1989-01-01

    The fatigue response of an Al-Li-Cu-Mg-Zr (8090) alloy has been studied at room temperature. The initiation and growth of small and long cracks has been examined at R = 0.1 and at a frequency of 100 Hz. Initiation was observed to occur dominantly at sub-grain boundaries. The growth of the small cracks was crystallographic in character and exhibited little evidence of retardation or arrest at the grain boundaries. The long crack data showed the alloy to have a high resistance to fatigue crack growth with underaging providing the optimum heat treatment for fatigue crack growth resistance. In general, this can be attributed to high levels of crack closure which resulted from the presence of extensive microstructurally related asperities.

  10. Application of spark plasma sintering (SPS) for the fabrication of in situ Ni–TiC nanocomposite clad layer

    International Nuclear Information System (INIS)

    Highlights: • Direct and indirect mechanical alloying was applied to fabricate a Ni–Ti–C metastable powder. • The metastable mechanically alloyed powder could undergo a reaction to synthesize TiC at high temperatures. • Spark plasma sintering was proposed as a cladding method to fabricate in situ Ni–TiC composite layer on steel. - Abstract: Spark plasma sintering (SPS) was utilized to create in situ Ni–TiC nanocomposite layers on steel substrates using reactive powders. Reactive Ni–Ti–C powders with a nominal composition of Ni–40 wt.%TiC were prepared by mechanical alloying (MA) in a high energy planetary ball mill. Two approaches were applied to prepare reactive powders. In the first approach a single-step method was conducted by milling Ni, Ti and C simultaneously. The second approach involved double step MA in which Ni–Ti and Ni–C powder mixtures were milled separately in the first step and the resultant powders were mixed and re-milled in the second step. Reactive powders were sintered successfully on St37 steel substrates by spark plasma sintering technique. X-ray diffractometery (XRD) was used to study the structural evolution during milling and after sintering. Powder particles and clad layers were examined by scanning electron microscopy (SEM) for microstructural investigations. Hardness measurements were conducted on the cross section of powder particles and clad layers. Sintering of reactive powders led to the in situ formation of TiC nano particles within the Ni matrix. Also the hardness of SPSed clad layers obtained from single step and double step MAed powders increased to 1250 and 780 HV, respectively. Clad layers showed a defect free interface with the steel substrate

  11. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  12. Spatial Mode Selective Waveguide with Hyperbolic Cladding

    CERN Document Server

    Tang, Y; Xu, M; Bäumer, S; Adam, A J L; Urbach, H P

    2016-01-01

    Hyperbolic Meta-Materials~(HMMs) are anisotropic materials with permittivity tensor that has both positive and negative eigenvalues. Here we report that by using a type II HMM as cladding material, a waveguide which only supports higher order modes can be achieved, while the lower order modes become leaky and are absorbed in the HMM cladding. This counter intuitive property can lead to novel application in optical communication and photonic integrated circuit. The loss in our HMM-Insulator-HMM~(HIH) waveguide is smaller than that of similar guided mode in a Metal-Insulator-Metal~(MIM) waveguide.

  13. Analysis of the behaviour of under-clad and surface cracks in cladded components

    International Nuclear Information System (INIS)

    The issue of the contribution is the characterization of under-clad and surface crack behaviour in ferritic steel components with an austenitic welded cladding. The experimental investigations were performed using large-scale samples. The residual stress field was determined in detail by a numerical simulation of the welding and heat treatment processes. These results were used for the numerical simulation of crack initiation and crack arrest. In all evaluated cases the crack was initiated in the ferritic material, while the cladding stayed intact even in case of a crack jump in the base metal. In the frame of case studies the results were transferred to application relevant geometries

  14. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  15. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V.

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  16. The Absorption Characteristics of Inhomogeneous Double-Clad Fibers

    Institute of Scientific and Technical Information of China (English)

    Hui Zhang; Zihua Wang; Zhongyin Xiao

    2003-01-01

    The absorption characteristics of radially inhomogeneous double-clad fiber (DCF) are investigated firstly with the method of caustic radius, combined with the method of WKBJ. The results are significant for double-clad optical fiber lasers and amplifiers.

  17. The Tests of Water Chemistry Effect on Corrosion Behaviour of Cladding

    International Nuclear Information System (INIS)

    To investigate the effect of improved primary water chemistry of PWR on corrosion behaviour of cladding, a project on compatibility of high content water chemistry and advanced cladding materials is in progress in NPIC during recent years. The modified Zr-4 and M5 alloys were tested in water with lithium at 3.50-6.48mg/L and boron at 983-3000mg/L for 3000 hours. For simulation of water-vapor environment around the surface of cladding, electrical heating elements were manufactured from modified Zr-4 and M5 cladding tubes. These elements were tested in water loops at 310oC and 15.5MPa with surface temperature of 346oC. The test results revealed that high concentration of lithium at same pH will accelerate the general corrosion rate and the hydrogen pick-up to a certain extent. Variation of pH300oC from 6.5 to 7.2 had no obvious influence on corrosion rate at 3.50 mg/l lithium. But the weight loss occurred when the boric acid content decreased to 6.48mg/l lithium. The weight gain difference between M5 and modified Zr-4 in single water phase was imperceptible. The thicker oxide layer with heavy deposits was formed on the boiling section of M5 elements; however the hydrogen absorption did not exceed that of modified Zr-4. It may imply that the M5 alloy is more sensitive to 'oxidation introduced deposition' under sub-cooled nucleate boiling condition, especially when oxygen is saturated. (author)

  18. AI-Li/SiCp composites and Ti-AI alloy powders and coatings prepared by a plasma spray atomization (PSA) technique

    Science.gov (United States)

    Khor, K. A.; Boey, F. Y. C.; Murakoshi, Y.; Sano, T.

    1994-06-01

    There has been increasing use of Al-Li alloys in the aerospace industry, due mainly to the low density and high elastic modulus of this material. However, the problem of low ductility and fracture toughness of this material has limited its present application to only weight- and stiffness-critical components. Development of Al-Li/ceramic composites is currently being investigated to enhance the service capabilities of this material. The Ti-Al alloy is also of interest to aerospace-type applications, engine components in particular, due to its attractive high-temperature properties. Preparation of fine powders by plasma melting of composite feedstock and coatings formed by plasma spraying was carried out to examine the effect of spray parameters on the microstructure and properties of these materials. Characterization of the powders and coatings was performed using the scanning electron microscope and image analyzer. Examination of the plasma-sprayed powders and coatings has shown that in the Al-Li/SiC composite there is melting of both materials to form a single composite particle. The SiC reinforcement was in the submicron range and contributed to additional strengthening of the composite body, which was formed by a cold isostatic press and consolidated by hot extrusion or hot forging processes. The plasma-sprayed Ti-Al powder showed four categories of microstructures: featureless, dendritic, cellular, and martensite-like.

  19. Thermal Shock Properties of Cladding with SiC{sub f}/SiC Composite Protective Films

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Donghee; Park, Kwangheon [Kyunghee University, Yongin (Korea, Republic of); Kim, Weonju; Park, Jiyeon; Kim, Daejong; Lee, Hyeon Geun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In general, Zr-4 alloy is used for such nuclear fuel cladding. Zr-4 possesses a very small thermal neutron absorption cross-section and has superior corrosion resistance in the normal operating conditions of a nuclear reactor. However, in the case of a critical accident such as a LOCA (loss-of-coolant accident) in the Fukushima disaster, the risk of hydrogen explosion becomes serious. That is, in the case of coolant leakage, a dramatic reaction between the nuclear fuel cladding and steam can cause a heating reaction accompanied by rapid high-temperature oxidation, while creating a huge amount of hydrogen. Hence, the search for an alternative material for nuclear fuel cladding is being actively undertaken. Ceramic-based nuclear fuel cladding is receiving much attention as a means of improving safety. SiC has excellent properties of resistance to high temperature and high exposure and superior mechanical properties, as well as a very small thermal neutron absorption cross-section (0.09 barns), which causes almost no decrease in mechanical strength or volume change following exposure. This experiment examined the thermal shock properties and microstructure of cladding that has SiCf/SiC composite protective film, using polycarbosilane preceramic polymer.

  20. Evaluation of residual stress near the weld overlay cladding by welding and post-weld heat treatment

    International Nuclear Information System (INIS)

    Austenitic stainless steel is welded as a cladding on the inner surface of a reactor pressure vessel (RPV) made of low alloy steel. In order to assess the structural integrity of the RPV precisely, the residual stress distribution caused by weld-overlay cladding and post-weld heat treatment (PWHT) is evaluated. Since the cladding layer is very thin compared to vessel wall, it is necessary to evaluate the residual stress distribution around the weld fusion line can be very steep. In this study, cladded specimens were fabricated using different welding methods. Residual stress measurements using both sectioning and deep hole drilling (DHD) methods were then performed to evaluate the residual stress distributions through the weld fusion line. Three-dimensional thermal-elastic-plastic-creep analyses based on finite element method were also conducted to evaluate the residual stress caused by weld-overlay cladding and PWHT. It was shown that analytical results provided reasonable agreements on weld residual stress with experimental results. It was also clarified that the main cause of residual stress due to welding and PWHT was the difference of thermal expansion between weld and base metals. (author)

  1. Structure and Properties of Ti-Nb-C Coatings Obtained by Non-vacuum Electron Beam Cladding

    Science.gov (United States)

    Lenivtseva, O. G.; Polyakov, I. A.; Lazurenko, D. V.; Lozhkin, V. S.

    2015-10-01

    In this study the structure and properties of surface-alloyed cp-titanium layers obtained by non-vacuum electron beam cladding of niobium carbide powders were analyzed. A thickness of coatings fabricated by single-layer cladding was 1.3 mm. Cladding of the second layer led to an increase in the thickness by 0.8 mm. It was found that titanium carbide particles of different morphology acted as strengthening structural elements. The X-ray diffraction (XRD) analysis revealed the presence of α-Ti (α'-Ti), β-Ti, and TiC in the cladded layer. The results of the energy dispersive X-ray (EDX) analysis indicated the presence of Nb in the titanium matrix as well as in the carbide phase. However, such phases as NbC and (Nb, Ti)C were not identified by the XRD analysis. Transmission electron microscopy (TEM) revealed zones containing an increased amount of Nb. The structure of these zones was represented by the β-Ti and ω-Ti precipitation. An average microhardness value of cladded layers was approximately 330 HV.

  2. Effects of Ce-rich RE additions and heat treatment on the microstructure and tensile properties of Mg-Li-Al-Zn-based alloy

    International Nuclear Information System (INIS)

    Research highlights: → Adding Ce-rich RE leads to the formation of Al2RE/Al3RE phases in LAZ532 alloy. → Ce-rich RE additions improve the tensile properties of the alloy. → The tensile properties of alloys processed by heat treatment are increased greatly. → Adding RE changes the fracture pattern of the alloy. - Abstract: As-cast Mg-5Li-3Al-2Zn-xRE (x = 0-2.5 wt.%) alloys were prepared under the ambient of pure argon, and the effects of Ce-rich rare earths (RE) and heat treatment on the microstructure and mechanical properties of Mg-Li-Al-Zn-based alloy were investigated. The results show that the main phase compositions of Mg-5Li-3Al-2Zn (LAZ532) alloy consist of α-Mg and AlLi. With the addition of RE, Al-RE precipitate forms, and increases gradually, whereas AlLi phase decreases. The room temperature tensile test reveals that the addition of RE could clearly improves the mechanical properties of alloys which are further improved after heat treatment. In more detail, excellent tensile strength and ductility are obtained in 1.5 wt.% RE containing alloy in as-cast state. After heat treatment, the 1.0 wt.% RE containing alloy attains superior tensile strength. The differences in tensile strength are related to the morphology, distribution of second phases and solid-solution strengthening in different alloy systems. In addition, the fracture pattern of the alloy is predominantly brittle cleavage and tends to be quasi-cleavage with RE addition.

  3. Comparison of two analytical methods for the local quantitative determination of lithium and boron contents in cladding materials

    International Nuclear Information System (INIS)

    Pressurized water reactors contain boric acid for reactivity control. As the acidic coolant conditions result in an increased attack of the circuit materials, LiOH is added to render the coolant slightly alkaline. However, LiOH can affect corrosion of the Zr alloy cladding. Thus the Li content in the oxide layers of irradiated fuel rods is of high interest, especially for new alloys (pathfinder rods). At the 'Paul Scherrer Institut' the lithium as well as the boron content in the oxide layers of claddings are determined by Secondary Ion Mass Spectrometry (SIMS). Quantification is performed by direct comparison with a Zircaloy-oxide layer implanted with B and Li. A new and independent method using Laser Ablation Inductively Coupled Plasma Mass Spectrometry was applied to cross-check the SIMS data. (authors)

  4. Pellet-clad mechanical interactions: Pellet-clad bond failure and strain relief

    International Nuclear Information System (INIS)

    The effects of pellet-clad mechanical interaction would be expected to be particularly severe in the presence of bonding between the fuel and the cladding. However, such bonding is observed far more frequently than is corresponding cladding damage. It has recently been shown that the radial stress in the bond during power changes is very large and tensile, and thus likely to cause failure of the bond. In this paper the likely azimuthal extent of this de-bonding is considered, and the relief of hoop stress which this offers is assessed. It is shown that the magnitude of this relief is such as to provide an explanation of the low cladding failure rate observed. (orig.)

  5. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  6. The measurement of residual stresses in claddings

    International Nuclear Information System (INIS)

    The ring core method, a variation of the hole drilling method for the measurement of biaxial residual stresses, has been extended to measure stresses from depths of about 5 to 25mm. It is now possible to measure the stress profiles of clad material. Examples of measured stress profiles are shown and compared with those obtained with a sectioning technique. (author)

  7. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    Science.gov (United States)

    Li, Bo-Shiuan

    Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

  8. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  9. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  10. Evolución de la fricción interna del material compuesto de matriz Al-Li 8090 reforzado con partículas de SiC

    Directory of Open Access Journals (Sweden)

    Gutiérrez-Urrutia, I.

    2001-04-01

    Full Text Available The present study has been undertaken to investigate the mechanism of thermal stress relief at the range of temperatures below room temperature for the metal matrix composite Al-Li 8090/SiC. For this aim the experimental technique of internal friction has been used which has been showed up very effective. Several thermal cycles from 453 K to 100 K were used in order to measure the internal friction as well as the elastic modulus of the material concluding that thermal stresses are relaxed by microplastic deformation around the reinforcements. It has been also related the variation in the elastic modulus with the different levels of precipitation.

    El presente trabajo investiga el mecanismo de relajación de tensiones térmicas a temperaturas por debajo de la de ambiente en el material compuesto Al-Li 8090/SiC. Para ello se ha empleado la técnica experimental de fricción interna que se ha mostrado la más eficaz para tal fin. Aplicando diferentes ciclos térmicos de 453 K a 100 K se midió tanto la fricción interna como el módulo elástico del material concluyendo que el mecanismo de relajación de tensiones térmicas es el de microdeformación plástica alrededor del reforzamiento. También se relaciona la variación del módulo elástico con los diferentes estadios de precipitación.

  11. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N.; Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Terrani, Kurt A.

    2015-11-01

    The high resistance of cladding to plastic deformation and burst failure is one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) since the deformation and burst behavior governs the cooling efficiency of flow channels and process of fission product release. To simulate and evaluate such deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisted of a high-resolution video camera, light filtering unit, and monochromatic light sources, and the in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. In this study eleven (11) candidate cladding materials for ATF, i.e., 6 FeCrAl alloys and 5 nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800°C while negligible strain rates were measured for higher strength alloys and/or for relatively thick wall specimens.

  12. Development of weldable, corrosion-resistant iron-aluminide alloys

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.; Goodwin, G.M.; Wang, X.L. [Oak Ridge National Laboratory, TN (United States)

    1995-05-01

    Corrosion-resistant, weldable FeAl alloys have been developed with improved high-temperature strength industrial applications. Previous processing difficulties with these alloys led to their evaluation as weld-overlay claddings on conventional structural steels to take advantage of their good properties now. Simplified and better processing methods for monolithic FeAl components are also currently being developed so that components for industrial testing can be made. Other avenues for producing FeAl coatings are currently being explored. Neutron scattering experiments residual stress distributions in the FeAl weld-overlay cladding began in FY 1993 and continued this year.

  13. Role of the micro/macro structure of welds in crack nucleation and propagation in aerospace aluminum-lithium alloy

    Science.gov (United States)

    Talia, George E.

    1996-01-01

    Al-Li alloys offer the benefits of increased strength, elastic modulus and lower densities as compared to conventional aluminum alloys. Martin Marietta Laboratories has developed an Al-Li alloy designated 2195 which is designated for use in the cryogenic tanks of the space shuttle. The Variable Polarity Plasma Arc (VPPA) welding process is currently being used to produce these welds [1]. VPPA welding utilizes high temperature ionized gas (plasma) to transfer heat to the workpiece. An inert gas, such as Helium, is used to shield the active welding zone to prevent contamination of the molten base metal with surrounding reactive atmospheric gases. [1] In the Space Shuttle application, two passes of the arc are used to complete a butt-type weld. The pressure of the plasma stream is increased during the first pass to force the arc entirely through the material, a practice commonly referred to as keyholing. Molten metal forms on either side of the arc and surface tension draws this liquid together as the arc passes. 2319 Al alloy filler material may also be fed into the weld zone during this pass. During the second pass, the plasma stream pressure is reduced such that only partial penetration of the base material is obtained. Al 2319 filler material is added during this pass to yield a uniform, fully filled welded joint. This additional pass also acts to alter the grain structure of the weld zone to yield a higher strength joint.

  14. Suitability of maraging steel weld cladding for repair of die casting tooling Part II

    OpenAIRE

    Taljat, Boštjan; Klobčar, Damjan; Muhič, Mitja; Tušek, Janez; Kosec, Ladislav

    2015-01-01

    This study was done to evaluate precipitation annealing of 18% Ni maraging steel repair welds during aluminium alloy die casting and to predict the prolonged in-service tool life. The emphasis of this study was the influence of post-weld precipitation annealing heat treatment and aluminium die casting thermal cycling on metallurgical and mechanical properties. A series of specimens of 1.2344 tool steel was prepared to which 1.6356 maraging steel wasgas tungsten arc weld clad. Analysis of weld...

  15. Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

    International Nuclear Information System (INIS)

    This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

  16. Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-23

    This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

  17. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

  18. Corrosion behaviour of 8090 alloy in saline solution with moderate aggressiveness

    International Nuclear Information System (INIS)

    Corrosion studies of Al-Li alloys are not so extensive and concentrate almost exclusively on atmospheric exposure tests and accelerated laboratory tests due to the fact they provide a reasonable approximation to the real behaviour of the alloy in service conditions. This paper attempts to establish a correlation between the evolution of the impedance diagrams and the process of the attack undergone by a commercial 8090 T8171 alloy, with the aim of establishing the kinetics of the corrosion process. After 100 h of immersion, samples showed only a slight intergranular attack. As a results of the low aggressiveness of the solution no major deviations from the ideal behaviour described by the Randles circuit are expected in the impedance plots. After 50 hours of testing, the impedance diagram evolves towards two semicircles which seem to be related with the charge transfer and ionic migration through the oxide layer and the adsorption of electrolyte anions. (Author) 7 refs

  19. Osteoblast interaction with laser cladded HA and SiO2-HA coatings on Ti-6Al-4V

    International Nuclear Information System (INIS)

    In order to improve the bioactivity and biocompatibility of titanium endosseous implants, the morphology and composition of the surfaces were modified. Polished Ti-6Al-4V substrates were coated by a laser cladding process with different precursors: 100 wt.% HA and 25 wt.% SiO2-HA. X-ray diffraction of the laser processed samples showed the presence of CaTiO3, Ca3(PO4)2, and Ca2SiO4 phases within the coatings. From in vitro studies, it was observed that compared to the unmodified substrate all laser cladded samples presented improved cellular interactions and bioactivity. The samples processed with 25 wt.% SiO2-HA precursor showed a significantly higher HA precipitation after immersion in simulated body fluid than 100 wt.% HA precursor and titanium substrates. The in vitro biocompatibility of the laser cladded coatings and titanium substrate was investigated by culturing of mouse MC3T3-E1 pre-osteoblast cell line and analyzing the cell viability, cell proliferation, and cell morphology. A significantly higher cell attachment and proliferation rate were observed for both laser cladded 100 wt.% HA and 25 wt.% SiO2-HA samples. Compared to 100 wt.% HA sample, 25 wt.% SiO2-HA samples presented a slightly improved cellular interaction due to the addition of SiO2. The staining of the actin filaments showed that the laser cladded samples induced a normal cytoskeleton and well-developed focal adhesion contacts. Scanning electron microscopic image of the cell cultured samples revealed better cell attachment and spreading for 25 wt.% SiO2-HA and 100 wt.% HA coatings than titanium substrate. These results suggest that the laser cladding process improves the bioactivity and biocompatibility of titanium. The observed biological improvements are mainly due to the coating induced changes in surface chemistry and surface morphology. Highlights: → Laser cladding of Ti alloys with bioceramics creates new phases. → Laser cladded samples with SiO2-doped bioceramics show higher

  20. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  1. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.

    2014-06-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  2. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  3. Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

    International Nuclear Information System (INIS)

    High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture toughness of high-burnup cladding is assumed to be sensitive to temperature and exhibit ductile-brittle transition phenomena similar to those of irradiated bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the material's transition temperature. Temperature dependence of fracture toughness is, in turn, sensitive to cladding microstructure such as density, distribution, and orientation of hydrides, oxygen distribution in the metallic phase, and irradiation-induced damage. Because all these factors are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, fuel failure is predicted to be strongly dependent on cladding axial location as well as on burnup. 10 figs, 21 refs

  4. Fatigue-crack propagation in advanced aerospace materials: Aluminum-lithium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Venkateswara Rao, K.T.; Ritchie, R.O.

    1988-10-01

    Characteristics of fatigue-crack propagation behavior are reviewed for recently developed commercial aluminum-lithium alloys, with emphasis on the underlying micromechanisms associated with crack advance and their implications to damage-tolerant design. Specifically, crack-growth kinetics in Alcoa 2090-T8E41, Alcan 8090 and 8091, and Pechiney 2091 alloys, and in certain powder-metallurgy alloys, are examined as a function of microstructure, plate orientation, temperature, crack size, load ratio and loading sequence. In general, it is found that growth rates for long (> 10 mm) cracks are nearly 2--3 orders of magnitude slower than in traditional 2000 and 7000 series alloys at comparable stress-intensity levels. In additions, Al-Li alloys shown enhanced crack-growth retardations following the application of tensile overloads and retain superior fatigue properties even after prolonged exposure at overaging temperatures; however, they are less impressive in the presence of compression overloads and further show accelerated crack-growth behavior for microstructurally-small (2--1000 {mu}m) cracks (some three orders of magnitude faster than long cracks). These contrasting observations are attributed to a very prominent role of crack-tip shielding during fatigue-crack growth in Al-Li alloys, promoted largely by the tortuous and zig-zag nature of the crack-path morphologies. Such crack paths result in locally reduced crack-tip stress intensities, due to crack deflection and consequent crack wedging from fracture-surface asperities (roughness-induced crack closure); however, such mechanisms are far less potent in the presence of compressive loads, which act to crush the asperities, and for small cracks, where the limited crack wake severely restricts the shielding effect. 50 refs., 21 figs.

  5. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    Science.gov (United States)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

  6. Conditioning of nuclear cladding wastes by melting

    International Nuclear Information System (INIS)

    This paper discusses a cold-crucible induction melting process to condition cladding waste from irradiated fast breeder reactor fuel. The process has been developed by the CEA at Marcoule (France) as part of a major R and D program. It has been qualified at industrial scale on nonradioactive waste, and at laboratory scale on radioactive waste: several radioactive ingots have been produced from actual stainless steel or zircaloy hulls. The results confirm the numerous advantages of this containment method

  7. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt% gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes, i.e., (1) partial amorphization of large Zr-Fe-Cr and Zr-Fe-Ni precipitates (300 to 800 nm in size), (2) virtually complete amorphization of small intermetallic precipitates and subsequent dissolution of the alloying elements, and (3) spinodal-like fluctuation and redistribution of the alloying elements following the amorphization and dissolution. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conducive to nodular oxidation. Besides the microstructural features associated with the intermetallic precipitates, TEM stereo electron microscopy revealed microscopic zirconium hydrides (30 to 100 nm in size) that were too small to be resolved by OM or SEM. Stereoscopic examination revealed a tendency for precipitation of the microscopic hydrides with c-component dislocations as the burnup increased. Also, an examination of bright- and dark-field stereopair images revealed three-dimensional distributions of fine cubic-zirconium-oxide precipitates (5 to 10 nm in size) and unidentified ''black-dot'' (5 to 10 nm) and ''white

  8. Study on high-performance fuel cladding materials. Joint research report in FY2001-2005. Phase 2 (Joint research)

    International Nuclear Information System (INIS)

    The research concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100GWd/t of BWR was pursued for 5 years from 2001 to 2005. Phase 2 for study on high-performance fuel cladding materials was planned as a joint research, by considering the effective use of MOX, minimizing both the electrical cost and radioactive waste. Comparing with UO2 with the maximum enrichment of 5% 235U, the advantage of MOX fuels is easy to achieve the high burnup, because it has the high enrichment factor of Puf up to 20% as same as it of LMFBR. On the Phase 1, the modified stainless steel of Fe-25Cr-35Ni-0.2Ti as fuel claddings and Nb-Mo alloy as a liner for inhibiting the pellet-clad interaction were selected as candidate materials, by evaluating fundamental properties required to BWR cladding materials, that are the nuclear economy, radioactivity, mass-transfer, irradiation properties, mechanical properties so on. On the present study, the making process of cladding tubes, lining by diffusion bonding, end plug by laser welding were developed and optimized, by considering the practical use of fuel elements consists of these candidates. The practical applicability was basically examined by irradiation tests using the accelerator of TIARA and the research reactor of JRR-3, for mainly confirming the resistance to IGSCC as one of the current important issues of BWR core materials of low carbon grade stainless steels. Creep and fatigue testing data were also obtained for evaluating the long performance of candidate materials. The behavior as fuel elements was analyzed with the safety calculation code for BWRs. The obtained results were established as a data base system, by considering the applicability to the fuel design and in-pile loop tests. (author)

  9. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    International Nuclear Information System (INIS)

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective

  10. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  11. New zirconium alloys for nuclear application

    International Nuclear Information System (INIS)

    Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions (∼ 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

  12. Pellet-Clad Mechanical Interaction Analysis with ANSYS Mechanical Module

    International Nuclear Information System (INIS)

    Pellet-Clad Mechanical Interaction (PCMI) has been known as a potential threat in fuel cladding integrity during power ramp conditions and high burn-up scenario. As the fuel outer surface contact with clad inner surface, the local stress become increased. Moreover, fuel pellet have much higher temperature in operation and have much greater expansion effects than clad, which occur additional contact pressure on clad inner surface, the cladding pellet deforms into a shape reflecting that of the pellet. This mechanical interaction between fuel pellet and clad depends on gap size, burn-up, friction coefficient between clad and pellet. Moreover, recent field result shows that nearly PCI-induced failures are thought to have developed at a missing pellet surface (MPS), where the tangential stress has its maximum and the cladding temperature has its minimum. For the additional study on PCMI, it is very important and valuable to find geometric parameters of MPS which make critical safety issue on cladding material safety. Followings are result and conclusion of the parametric studies

  13. Cladding failure by local plastic instability

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, J.M.; Deitrich, L.W.

    1977-12-01

    Cladding failure is one of the major considerations in analysis of fuel-pin behavior during hypothetical accident transients since time, location, and nature of failure govern the early postfailure material motion and reactivity feedback. Out-of-pile thermal transient tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture and that the extent of local instability limits the initial rip length. To investigate the details of bulge formation and growth, a perturbation analysis of the equations governing large deformation of a cylindrical shell has been developed, resulting in a set of linear differential equations for the bulge geometry. These equations have been solved along with appropriate constitutive equations and various constraints on the ends of the cladding. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or nonsymmetric cooling. Of these, only the first is relevant to out-of-pile burst tests. Here it has been found that the most likely imperfection that will grow unstably to failure leads to a bulge around half the circumference with an axial length 1.1 times the deformed diameter. This is in general agreement with burst-test results. For the case of in-reactor fuel pins, it has been found that thermal perturbations can significantly affect local instability, particularly if the deformation process is thermally activated with a high activation energy.

  14. Influence of laser cladding regimes on structural features and mechanical properties of coatings on titanium substrates

    Science.gov (United States)

    Malyutina, Yulia N.; Lazurenko, Daria V.; Bataev, Ivan A.; Movtchan, Igor A.

    2015-10-01

    In this paper an influence of the tantalum content on the structure and properties of surface layers of the titanium alloy doped using a laser treatment technology was investigated. It was found that an increase of a quantity of filler powder per one millimeter of a track length contributed to a rise of the content of undissolved particles in coatings. The maximum thickness of a cladded layer was reached at the mass of powder per the length unit equaled to 5.5 g/cm. Coatings were characterized by the formation of a dendrite structure with attributes of segregation. The width of a quenched fusion zone grew with an increase in the rate of powder feed to the treated area. Significant strengthening of the titanium surface layer alloyed with tantalum was not observed; however, the presence of undissolved tantalum particles can decrease the hardness of titanium surface layers.

  15. Thermodynamic studies of thorium carbide fuel preparation and fuel/clad compatibility

    International Nuclear Information System (INIS)

    The carbothermic reduction of thorium and uranium-thorium dioxide to monocarbide has been assessed. Equilibrium calculations have yielded Th-C-O and U-Th-C-O phase equilibria and CO pressures generated during reduction. The CO pressures were found to be at least five orders of magnitude greater than any of the other 15 gaseous species considered. This confirms that the monocarbide can successfully be prepared by carbothermic reduction. The chemical compatibility of thorium carbides with the Cr-Fe-Ni content of clad alloys has been thermodynamically evaluated. Solid solutions of 5> and 5> and of 7C3> and 7C3> were the principal reaction products. The Cr-Fe-Ni content of 316 stainless steel showed much less reaction product than that of any of the other six alloys considered. (author)

  16. Microstructure stability of candidate stainless steels for Gen-IV SCWR fuel cladding application

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jian, E-mail: jili@nrcan.gc.ca [CanmetMATERIALS, Natural Resources Canada, 183 Longwood Rd. S., Hamilton, ON (Canada); Zheng, W. [CanmetMATERIALS, Natural Resources Canada, 183 Longwood Rd. S., Hamilton, ON (Canada); Penttilä, S. [VTT Technical Research Center of Finland, Materials for Power Engineering, P.O. Box 1000, FI-02044 VTT (Finland); Liu, P. [CanmetMATERIALS, Natural Resources Canada, 183 Longwood Rd. S., Hamilton, ON (Canada); Woo, O.T.; Guzonas, D. [AECL Chalk River Laboratory, Chalk River, Ontario (Canada)

    2014-11-15

    In the past few years, significant progress has been made in materials selection for Gen-IV SCWR fuel cladding applications. Current studies indicate that austenite stainless steels such as 310H are promising candidates for in-core applications. Alloys in this group are promising for their corrosion resistance, SCC resistance, high temperature mechanical properties and creep resistance at temperatures up to 700 °C. However, one under-studied area of this alloy is the long-term microstructure stability under the proposed reactor operating condition. Unstable microstructure not only results in embrittlement but also has the potential to reduce their resistance to corrosion or stress-corrosion cracking. In this study, stainless steels 310H and 304H were tested for their SCWR corrosion resistance and microstructure stability.

  17. Influence of laser cladding regimes on structural features and mechanical properties of coatings on titanium substrates

    Energy Technology Data Exchange (ETDEWEB)

    Malyutina, Yulia N., E-mail: iuliiamaliutina@gmail.ru; Lazurenko, Daria V., E-mail: pavlyukova-87@mail.ru; Bataev, Ivan A., E-mail: ivanbataev@ngs.ru [Novosibirsk State Technical University, Novosibirsk, 630073 (Russian Federation); Movtchan, Igor A., E-mail: igor.movtchan@enise.fr [National Engineering School in Saint-Etienne, Saint-Etienne, 42000 France (France)

    2015-10-27

    In this paper an influence of the tantalum content on the structure and properties of surface layers of the titanium alloy doped using a laser treatment technology was investigated. It was found that an increase of a quantity of filler powder per one millimeter of a track length contributed to a rise of the content of undissolved particles in coatings. The maximum thickness of a cladded layer was reached at the mass of powder per the length unit equaled to 5.5 g/cm. Coatings were characterized by the formation of a dendrite structure with attributes of segregation. The width of a quenched fusion zone grew with an increase in the rate of powder feed to the treated area. Significant strengthening of the titanium surface layer alloyed with tantalum was not observed; however, the presence of undissolved tantalum particles can decrease the hardness of titanium surface layers.

  18. Influence of laser cladding regimes on structural features and mechanical properties of coatings on titanium substrates

    International Nuclear Information System (INIS)

    In this paper an influence of the tantalum content on the structure and properties of surface layers of the titanium alloy doped using a laser treatment technology was investigated. It was found that an increase of a quantity of filler powder per one millimeter of a track length contributed to a rise of the content of undissolved particles in coatings. The maximum thickness of a cladded layer was reached at the mass of powder per the length unit equaled to 5.5 g/cm. Coatings were characterized by the formation of a dendrite structure with attributes of segregation. The width of a quenched fusion zone grew with an increase in the rate of powder feed to the treated area. Significant strengthening of the titanium surface layer alloyed with tantalum was not observed; however, the presence of undissolved tantalum particles can decrease the hardness of titanium surface layers

  19. Ageing of zirconium alloy components

    International Nuclear Information System (INIS)

    India has two types (pressurized heavy water reactors (PHWRs) and boiling water reactors (BWRs)) of commercial nuclear reactors in operation, in addition to research reactors. Many of the life limiting critical components in these reactors are fabricated from zirconium alloys. The progressive degradation of these components caused by the cumulative exposure of high energy neutron irradiation with increasing period of reactor operation was monitored to assess the degree of ageing. The components/specimens examined included fuel element claddings removed from BWRs, pressure tubes and garter springs removed from PHWRs and calandria tube specimens used in PHWRs. The tests included tension test (for cladding, garter spring), fracture toughness test (for pressure tube), crush test (for garter spring), and measurement of irradiation induced growth (for calandria tube). Results of various tests conducted are presented and applications of the test results are elaborated for residual life estimation/life extension of the components

  20. Corrosion of aluminium-clad spent fuel in LVR-15 Research Reactor storage facilities

    International Nuclear Information System (INIS)

    This report documents the work performed under the IAEA Coordinated Research Project (CRP) on corrosion of the research reactor aluminium-clad spent fuel in water in the Nuclear Research Institute Rez. The aim of the project was to evaluate the corrosion of coupons of aluminium alloys used as cladding material of research reactor fuel elements, upon exposure to the water on the spent fuel storage basins. The corrosion of coupons exposed to two storage facilities at our Institute was investigated. Test racks were delivered by the IAEA and these contained coupons of two aluminium alloys, AA 6061 and SZAV-1. The racks also contained bimetallic couples consisting of aluminium alloy and stainless steel coupons. Rolled and extruded AA 6061 coupons were also tested. The single coupons, bimetallic couples and coupons with crevice couples were immersed in the at-reactor basin (ARB) and in the high-level waste pool (HLW). The chemical parameters of water in the two storage facilities were monitored and the extent of sedimentation of solids was measured. The ionic impurities were mainly Cl- and SO42- and their contents were 2 -15 μg/l in the HLW pool and about 20-250 μg/l in ARB. The iron content was below 2 μg/l in both facilities. After two years of exposure, pitting of the coupons was evaluated. Pits were observed mainly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between pitting and the type of aluminium alloy or between rolled and extruded materials. In the bimetallic couples, contact with stainless steel coupons did not have any affect on localized corrosion of the Al coupons. The pit depths were less than 50 μm on most of the coupon surfaces. Data obtained at this Institute should be compared with the results of other participants of this CRP. (author)